Sample records for safety research reactor

  1. Conversion Preliminary Safety Analysis Report for the NIST Research Reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Diamond, D. J.; Baek, J. S.; Hanson, A. L.

    The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the NIST research reactor (aka NBSR); a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in anmore » aluminum alloy, and the development of the fabrication techniques. This report is a preliminary version of the Safety Analysis Report (SAR) that would be submitted to the U.S. Nuclear Regulatory Commission (NRC) for approval prior to conversion. The report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis in any conversion SAR is to explain the differences between the LEU and HEU cores and to show the acceptability of the new design; there is no need to repeat information regarding the current reactor that will not change upon conversion. Hence, as seen in the report, the bulk of the SAR is devoted to Chapter 4, Reactor Description, and Chapter 13, Safety Analysis.« less

  2. A new safety channel based on ¹⁷N detection in research reactors.

    PubMed

    Seyfi, Somayye; Gharib, Morteza

    2015-10-01

    Tehran research reactor (TRR) is a representative of pool type research reactors using light water, as coolant and moderator. This reactor is chosen as a prototype to demonstrate and prove the feasibility of (17)N detection as a new redundant channel for reactor power measurement. In TRR, similar to other pool type reactors, neutron detectors are immersed in the pool around the core as the main power measuring devices. In the present article, a different approach, using out of water neutron detector, is employed to measure reactor power. This new method is based on (17)O (n,p) (17)N reaction taking place inside the core and subsequent measurement of delayed neutrons emitted due to (17)N disintegration. Count and measurement of neutrons around outlet water pipe provides a reliable redundant safety channel to measure reactor power. Results compared with other established channels indicate a good agreement and shows a linear interdependency with true thermal power. Safety of reactor operation is improved with installation & use of this new power measuring channel. The new approach may equally serve well as a redundant channel in all other types of reactors having coolant comprised of oxygen in its molecular constituents. Contrary to existing channels, this one is totally out of water and thus is an advantage over current instrumentations. It is proposed to employ the same idea on other reactors (nuclear power plants too) to improve safety criteria. Copyright © 2015 Elsevier Ltd. All rights reserved.

  3. Safety Issues at the DOE Test and Research Reactors. A Report to the U.S. Department of Energy.

    ERIC Educational Resources Information Center

    National Academy of Sciences - National Research Council, Washington, DC. Commission on Physical Sciences, Mathematics, and Resources.

    This report provides an assessment of safety issues at the Department of Energy (DOE) test and research reactors. Part A identifies six safety issues of the reactors. These issues include the safety design philosophy, the conduct of safety reviews, the performance of probabilistic risk assessments, the reliance on reactor operators, the fragmented…

  4. A probabilistic safety analysis of incidents in nuclear research reactors.

    PubMed

    Lopes, Valdir Maciel; Agostinho Angelo Sordi, Gian Maria; Moralles, Mauricio; Filho, Tufic Madi

    2012-06-01

    This work aims to evaluate the potential risks of incidents in nuclear research reactors. For its development, two databases of the International Atomic Energy Agency (IAEA) were used: the Research Reactor Data Base (RRDB) and the Incident Report System for Research Reactor (IRSRR). For this study, the probabilistic safety analysis (PSA) was used. To obtain the result of the probability calculations for PSA, the theory and equations in the paper IAEA TECDOC-636 were used. A specific program to analyse the probabilities was developed within the main program, Scilab 5.1.1. for two distributions, Fischer and chi-square, both with the confidence level of 90 %. Using Sordi equations, the maximum admissible doses to compare with the risk limits established by the International Commission on Radiological Protection (ICRP) were obtained. All results achieved with this probability analysis led to the conclusion that the incidents which occurred had radiation doses within the stochastic effects reference interval established by the ICRP-64.

  5. Nuclear reactor safety research since three mile island.

    PubMed

    Mynatt, F R

    1982-04-09

    The Three Mile Island nuclear power plant accident has resulted in redirection of reactor safety research priorities. The small release to the environment of radioactive iodine-13 to 17 curies in a total radioactivity release of 2.4 million to 13 million curies-has led to a new emphasis on the physical chemistry of fission product behavior in accidents; the fact that the nuclear core was severely damaged but did not melt down has opened a new accident regime-that of the degraded core; the role of the operators in the progression and severity of the accident has shifted emphasis from equipment reliability to human reliability. As research progresses in these areas, the technical base for regulation and risk analysis will change substantially.

  6. Light-Water-Reactor safety research program. Quarterly progress report, January--March 1977

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    None

    The report summarizes the Argonne National Laboratory work performed during January, February, and March 1977 on water-reactor-safety problems. The following research and development areas are covered: (1) loss-of-coolant accident research: heat transfer and fluid dynamics; (2) transient fuel response and fission-product release program; (3) mechanical properties of zircaloy containing oxygen; and (4) steam-explosion studies.

  7. Shielding calculation and criticality safety analysis of spent fuel transportation cask in research reactors.

    PubMed

    Mohammadi, A; Hassanzadeh, M; Gharib, M

    2016-02-01

    In this study, shielding calculation and criticality safety analysis were carried out for general material testing reactor (MTR) research reactors interim storage and relevant transportation cask. During these processes, three major terms were considered: source term, shielding, and criticality calculations. The Monte Carlo transport code MCNP5 was used for shielding calculation and criticality safety analysis and ORIGEN2.1 code for source term calculation. According to the results obtained, a cylindrical cask with body, top, and bottom thicknesses of 18, 13, and 13 cm, respectively, was accepted as the dual-purpose cask. Furthermore, it is shown that the total dose rates are below the normal transport criteria that meet the standards specified. Copyright © 2015 Elsevier Ltd. All rights reserved.

  8. International Research Reactor Decommissioning Project

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Leopando, Leonardo; Warnecke, Ernst

    2008-01-15

    Many research reactors have been or will be shut down and are candidates for decommissioning. Most of the respective countries neither have a decommissioning policy nor the required expertise and funds to effectively implement a decommissioning project. The IAEA established the Research Reactor Decommissioning Demonstration Project (R{sup 2}D{sup 2}P) to help answer this need. It was agreed to involve the Philippine Research Reactor (PRR-1) as model reactor to demonstrate 'hands-on' experience as it is just starting the decommissioning process. Other facilities may be included in the project as they fit into the scope of R{sup 2}D{sup 2}P and complement tomore » the PRR-1 decommissioning activities. The key outcome of the R{sup 2}D{sup 2}P will be the decommissioning of the PRR-1 reactor. On the way to this final goal the preparation of safety related documents (i.e., decommissioning plan, environmental impact assessment, safety analysis report, health and safety plan, cost estimate, etc.) and the licensing process as well as the actual dismantling activities could provide a model to other countries involved in the project. It is expected that the R{sup 2}D{sup 2}P would initiate activities related to planning and funding of decommissioning activities in the participating countries if that has not yet been done.« less

  9. Reactor safety method

    DOEpatents

    Vachon, Lawrence J.

    1980-03-11

    This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

  10. Neutronic safety parameters and transient analyses for Poland's MARIA research reactor.

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Bretscher, M. M.; Hanan, N. A.; Matos, J. E.

    1999-09-27

    Reactor kinetic parameters, reactivity feedback coefficients, and control rod reactivity worths have been calculated for the MARIA Research Reactor (Swierk, Poland) for M6-type fuel assemblies with {sup 235}U enrichments of 80% and 19.7%. Kinetic parameters were evaluated for family-dependent effective delayed neutron fractions, decay constants, and prompt neutron lifetimes and neutron generation times. Reactivity feedback coefficients were determined for fuel Doppler coefficients, coolant (H{sub 2}O) void and temperature coefficients, and for in-core and ex-core beryllium temperature coefficients. Total and differential control rod worths and safety rod worths were calculated for each fuel type. These parameters were used to calculate genericmore » transients for fast and slow reactivity insertions with both HEU and LEU fuels. The analyses show that the HEU and LEU cores have very similar responses to these transients.« less

  11. Methods and strategies for future reactor safety goals

    NASA Astrophysics Data System (ADS)

    Arndt, Steven Andrew

    -informed analyses and discussions. This dissertation examines potential approaches to updating the safety goals that include the establishment of new quantitative safety goal associated with the comparative risk of generating electricity by viable competing technologies and modifications of the goals to account for multi-plant reactor sites, and issues associated with the use of safety goals in both initial licensing and operational decision making. This research develops a new quantitative health objective that uses a comparable benefit risk metric based on the life-cycle risk of the construction, operation and decommissioning of a comparable non-nuclear electric generation facility, as well as the risks associated with mining and transportation. This dissertation also evaluates the effects of using various methods for aggregating site risk as a safety metric, as opposed to using single plant safety goals. Additionally, a number of important assumptions inherent in the current safety goals, including the effect of other potential negative societal effects such as the generation of greenhouse gases (e.g., carbon dioxide) have on the risk of electric power production and their effects on the setting of safety goals, is explored. Finally, the role risk perception should play in establishing safety goals has been explored. To complete this evaluation, a new method to analytically compare alternative technologies of generating electricity was developed, including development of a new way to evaluate risk perception, and a new method was developed for evaluating the risk at multiple units on a single site. To test these modifications to the safety goals a number of possible reactor designs and configurations were evaluated using these new proposed safety goals to determine the goals' usefulness and utility. The results of the analysis showed that the modifications provide measures that more closely evaluate the potential risk to the public from the operation of nuclear power plants than

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Monteleone, S.

    1993-10-01

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

  13. Virtual environments simulation in research reactor

    NASA Astrophysics Data System (ADS)

    Muhamad, Shalina Bt. Sheik; Bahrin, Muhammad Hannan Bin

    2017-01-01

    Virtual reality based simulations are interactive and engaging. It has the useful potential in improving safety training. Virtual reality technology can be used to train workers who are unfamiliar with the physical layout of an area. In this study, a simulation program based on the virtual environment at research reactor was developed. The platform used for virtual simulation is 3DVia software for which it's rendering capabilities, physics for movement and collision and interactive navigation features have been taken advantage of. A real research reactor was virtually modelled and simulated with the model of avatars adopted to simulate walking. Collision detection algorithms were developed for various parts of the 3D building and avatars to restrain the avatars to certain regions of the virtual environment. A user can control the avatar to move around inside the virtual environment. Thus, this work can assist in the training of personnel, as in evaluating the radiological safety of the research reactor facility.

  14. Safety control circuit for a neutronic reactor

    DOEpatents

    Ellsworth, Howard C.

    2004-04-27

    A neutronic reactor comprising an active portion containing material fissionable by neutrons of thermal energy, means to control a neutronic chain reaction within the reactor comprising a safety device and a regulating device, a safety device including means defining a vertical channel extending into the reactor from an aperture in the upper surface of the reactor, a rod containing neutron-absorbing materials slidably disposed within the channel, means for maintaining the safety rod in a withdrawn position relative to the active portion of the reactor including means for releasing said rod on actuation thereof, a hopper mounted above the active portion of the reactor having a door disposed at the bottom of the hopper opening into the vertical channel, a plurality of bodies of neutron-absorbing materials disposed within the hopper, and means responsive to the failure of the safety rod on actuation thereof to enter the active portion of the reactor for opening the door in the hopper.

  15. Safety philosophy of gas turbine high temperature reactor (GTHTR300)

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Shoji Katanishi; Kazuhiko Kunitomi; Shusaku Shiozawa

    2002-07-01

    Japan Atomic Energy Research Institute (JAERI) has undertaken the study of an original design concept of gas turbine high temperature reactor, the GTHTR300. The general concept of this study is development of a greatly simplified design that leads to substantially reduced technical and cost requirements. Newly proposed design features enable the GTHTR300 to be an efficient and economically competitive reactor in 2010's. Also, the GTHTR300 fully takes advantage of its inherent safety characteristics. The safety philosophy of the GTHTR300 is developed based on the HTTR (High Temperature Engineering Test Reactor) of JAERI which is the first HTGR in Japan. Majormore » features of the newly proposed safety philosophy for the GTHTR300 are described in this article. (authors)« less

  16. New reactor technology: safety improvements in nuclear power systems.

    PubMed

    Corradini, M L

    2007-11-01

    Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems.

  17. Advanced reactors and associated fuel cycle facilities: safety and environmental impacts.

    PubMed

    Hill, R N; Nutt, W M; Laidler, J J

    2011-01-01

    The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described. Copyright © 2010 Health Physics Society

  18. Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Farmer, Mitchell T.; Bunt, R.; Corradini, M.

    The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy’s (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affectmore » reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).« less

  19. Nuclear reactor safety device

    DOEpatents

    Hutter, Ernest

    1986-01-01

    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.

  20. Request for Naval Reactors Comment on Proposed Prometheus Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to JPL

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    D. Kokkinos

    2005-04-28

    The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophymore » on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory.« less

  1. Nuclear reactor fuel containment safety structure

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Rosewell, M.P.

    A nuclear reactor fuel containment safety structure is disclosed and is shown to include an atomic reactor fuel shield with a fuel containment chamber and exhaust passage means, and a deactivating containment base attached beneath the fuel reactor shield and having exhaust passages, manifold, and fluxing and control material and vessels. 1 claim, 8 figures.

  2. Nuclear Reactor Safety--The APS Submits its Report

    ERIC Educational Resources Information Center

    Physics Today, 1975

    1975-01-01

    Presents the summary section of the American Physical Society (APS) report on the safety features of the light-water reactor, reviews the design, construction, and operation of a reactor and outlines the primary engineered safety features. Summarizes the major recommendations of the study group. (GS)

  3. Monitoring circuit for reactor safety systems

    DOEpatents

    Keefe, Donald J.

    1976-01-01

    The ratio between the output signals of a pair of reactor safety channels is monitored. When ratio falls outside of a predetermined range, it indicates that one or more of the safety channels has malfunctioned.

  4. Physics of reactor safety. Quarterly report, January--March 1977. [LMFBR

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    None

    1977-06-01

    This report summarizes work done on reactor safety, Monte Carlo analysis of safety-related critical assembly experiments, and planning of DEMI safety-related critical experiments. Work on reactor core thermal-hydraulics is also included.

  5. VVER Reactor Safety in Eastern Europe and Former Soviet Union

    NASA Astrophysics Data System (ADS)

    Papadopoulou, Demetra

    2012-02-01

    VVER Soviet-designed reactors that operate in Eastern Europe and former Soviet republics have heightened international concern for years due to major safety deficiencies. The governments of countries with VVER reactors have invested millions of dollars toward improving the safety of their nuclear power plants. Most of these reactors will continue to operate for the foreseeable future since they provide urgently-needed electrical power. Given this situation, this paper assesses the radiological consequences of a major nuclear accident in Eastern Europe. The paper also chronicles the efforts launched by the international nuclear community to improve the safety of the reactors and notes the progress made so far through extensive collaborative efforts in Armenia, Bulgaria, the Czech Republic, Hungary, Kazakhstan, Lithuania, Russia, Slovakia, and Ukraine to reduce the risks of nuclear accidents. Western scientific and technical staff collaborated with these countries to improve the safety of their reactor operations by strengthening the ability of the regulator to perform its oversight function, installing safety equipment and technologies, investing time in safety training, and working diligently to establish an enduring safety culture. Still, continued safety improvement efforts are necessary to ensure safe operating practices and achieve timely phase-out of older plants.

  6. Generation III reactors safety requirements and the design solutions

    NASA Astrophysics Data System (ADS)

    Felten, P.

    2009-03-01

    Nuclear energy's public acceptance, and hence its development, depends on its safety. As a reactor designer, we will first briefly remind the basic safety principles of nuclear reactors' design. We will then show how the industry, and in particular Areva with its EPR, made design evolution in the wake of the Three Miles Island accident in 1979. In particular, for this new generation of reactors, severe accidents are taken into account beyond the standard design basis accidents. Today, Areva's EPR meets all so-called "generation III" safety requirements and was licensed by several nuclear safety authorities in the world. Many innovative solutions are integrated in the EPR, some of which will be introduced here.

  7. ReactorHealth Physics operations at the NIST center for neutron research.

    PubMed

    Johnston, Thomas P

    2015-02-01

    Performing health physics and radiation safety functions under a special nuclear material license and a research and test reactor license at a major government research and development laboratory encompasses many elements not encountered by industrial, general, or broad scope licenses. This article reviews elements of the health physics and radiation safety program at the NIST Center for Neutron Research, including the early history and discovery of the neutron, applications of neutron research, reactor overview, safety and security of radiation sources and radioactive material, and general health physics procedures. These comprise precautions and control of tritium, training program, neutron beam sample processing, laboratory audits, inventory and leak tests, meter calibration, repair and evaluation, radioactive waste management, and emergency response. In addition, the radiation monitoring systems will be reviewed including confinement building monitoring, ventilation filter radiation monitors, secondary coolant monitors, gaseous fission product monitors, gas monitors, ventilation tritium monitor, and the plant effluent monitor systems.

  8. The Need for Cyber-Informed Engineering Expertise for Nuclear Research Reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Anderson, Robert Stephen

    Engineering disciplines may not currently understand or fully embrace cyber security aspects as they apply towards analysis, design, operation, and maintenance of nuclear research reactors. Research reactors include a wide range of diverse co-located facilities and designs necessary to meet specific operational research objectives. Because of the nature of research reactors (reduced thermal energy and fission product inventory), hazards and risks may not have received the same scrutiny as normally associated with power reactors. Similarly, security may not have been emphasized either. However, the lack of sound cybersecurity defenses may lead to both safety and security impacts. Risk management methodologiesmore » may not contain the foundational assumptions required to address the intelligent adversary’s capabilities in malevolent cyber attacks. Although most research reactors are old and may not have the same digital footprint as newer facilities, any digital instrument and control function must be considered as a potential attack platform that can lead to sabotage or theft of nuclear material, especially for some research reactors that store highly enriched uranium. This paper will provide a discussion about the need for cyber-informed engineering practices that include the entire engineering lifecycle. Cyber-informed engineering as referenced in this paper is the inclusion of cybersecurity aspects into the engineering process. A discussion will consider several attributes of this process evaluating the long-term goal of developing additional cyber safety basis analysis and trust principles. With a culture of free information sharing exchanges, and potentially a lack of security expertise, new risk analysis and design methodologies need to be developed to address this rapidly evolving (cyber) threatscape.« less

  9. Automatic safety rod for reactors

    DOEpatents

    Germer, John H.

    1988-01-01

    An automatic safety rod for a nuclear reactor containing neutron absorbing material and designed to be inserted into a reactor core after a loss-of-core flow. Actuation is based upon either a sudden decrease in core pressure drop or the pressure drop decreases below a predetermined minimum value. The automatic control rod includes a pressure regulating device whereby a controlled decrease in operating pressure due to reduced coolant flow does not cause the rod to drop into the core.

  10. Evaluation of nuclear-facility decommissioning projects. Summary report: Ames Laboratory Research Reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Link, B.W.; Miller, R.L.

    1983-07-01

    This document summarizes the available information concerning the decommissioning of the Ames Laboratory Research Reactor (ALRR), a five-megawatt heavy water moderated and cooled research reactor. The data were placed in a computerized information retrieval/manipulation system which permits its future utilization for purposes of comparative analysis. This information is presented both in detail in its computer output form and also as a manually assembled summarization which highlights the more important aspects of the decommissioning program. Some comparative information with reference to generic decommissioning data extracted from NUREG/CR 1756, Technology, Safety and Costs of Decommissioning Nuclear Research and Test Reactors, is included.

  11. An approach to model reactor core nodalization for deterministic safety analysis

    NASA Astrophysics Data System (ADS)

    Salim, Mohd Faiz; Samsudin, Mohd Rafie; Mamat @ Ibrahim, Mohd Rizal; Roslan, Ridha; Sadri, Abd Aziz; Farid, Mohd Fairus Abd

    2016-01-01

    Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH1.6, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

  12. An approach to model reactor core nodalization for deterministic safety analysis

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my; Samsudin, Mohd Rafie, E-mail: rafies@tnb.com.my; Mamat Ibrahim, Mohd Rizal, E-mail: m-rizal@nuclearmalaysia.gov.my

    Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to bemore » employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH{sub 1.6}, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D{sup ®} computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.« less

  13. Safety and Environment aspects of Tokamak- type Fusion Power Reactor- An Overview

    NASA Astrophysics Data System (ADS)

    Doshi, Bharat; Reddy, D. Chenna

    2017-04-01

    Naturally occurring thermonuclear fusion reaction (of light atoms to form a heavier nucleus) in the sun and every star in the universe, releases incredible amounts of energy. Demonstrating the controlled and sustained reaction of deuterium-tritium plasma should enable the development of fusion as an energy source here on Earth. The promising fusion power reactors could be operated on the deuterium-tritium fuel cycle with fuel self-sufficiency. The potential impact of fusion power on the environment and the possible risks associated with operating large-scale fusion power plants is being studied by different countries. The results show that fusion can be a very safe and sustainable energy source. A fusion power plant possesses not only intrinsic advantages with respect to safety compared to other sources of energy, but also a negligible long term impact on the environment provided certain precautions are taken in its design. One of the important considerations is in the selection of low activation structural materials for reactor vessel. Selection of the materials for first wall and breeding blanket components is also important from safety issues. It is possible to fully benefit from the advantages of fusion energy if safety and environmental concerns are taken into account when considering the conceptual studies of a reactor design. The significant safety hazards are due to the tritium inventory and energetic neutron fluence induced activity in the reactor vessel, first wall components, blanket system etc. The potential of release of radioactivity under operational and accident conditions needs attention while designing the fusion reactor. Appropriate safety analysis for the quantification of the risk shall be done following different methods such as FFMEA (Functional Failure Modes and Effects Analysis) and HAZOP (Hazards and operability). Level of safety and safety classification such as nuclear safety and non-nuclear safety is very important for the FPR (Fusion

  14. Automatic safety rod for reactors. [LMFBR

    DOEpatents

    Germer, J.H.

    1982-03-23

    An automatic safety rod for a nuclear reactor containing neutron absorbing material and designed to be inserted into a reactor core after a loss-of-flow. Actuation is based upon either a sudden decrease in core pressure drop or the pressure drop decreases below a predetermined minimum value. The automatic control rod includes a pressure regulating device whereby a controlled decrease in operating pressure due to reduced coolant flow does not cause the rod to drop into the core.

  15. Light Water Reactor Sustainability Program Reactor Safety Technologies Pathway Technical Program Plan

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Corradini, M. L.; Peko, D.; Farmer, M.

    In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safetymore » initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Corradini, M. L.

    In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safetymore » initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the

  17. Advanced Test Reactor Safety Basis Upgrade Lessons Learned Relative to Design Basis Verification and Safety Basis Management

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    G. L. Sharp; R. T. McCracken

    The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The reactor also provides other irradiation services such as radioisotope production. The ATR and its support facilities are located at the Test Reactor Area of the Idaho National Engineering and Environmental Laboratory (INEEL). An audit conducted by the Department of Energy's Office of Independent Oversight and Performance Assurance (DOE OA) raised concerns that design conditions at the ATR were not adequately analyzedmore » in the safety analysis and that legacy design basis management practices had the potential to further impact safe operation of the facility.1 The concerns identified by the audit team, and issues raised during additional reviews performed by ATR safety analysts, were evaluated through the unreviewed safety question process resulting in shutdown of the ATR for more than three months while these concerns were resolved. Past management of the ATR safety basis, relative to facility design basis management and change control, led to concerns that discrepancies in the safety basis may have developed. Although not required by DOE orders or regulations, not performing design basis verification in conjunction with development of the 10 CFR 830 Subpart B upgraded safety basis allowed these potential weaknesses to be carried forward. Configuration management and a clear definition of the existing facility design basis have a direct relation to developing and maintaining a high quality safety basis which properly identifies and mitigates all hazards and postulated accident conditions. These relations and the impact of past safety basis management practices have been reviewed in order to identify lessons learned from the safety basis upgrade process and appropriate actions to resolve possible concerns with respect to the current ATR

  18. The use of experimental data in an MTR-type nuclear reactor safety analysis

    NASA Astrophysics Data System (ADS)

    Day, Simon E.

    Reactivity initiated accidents (RIAs) are a category of events required for research reactor safety analysis. A subset of this is unprotected RIAs in which mechanical systems or human intervention are not credited in the response of the system. Light-water cooled and moderated MTR-type ( i.e., aluminum-clad uranium plate fuel) reactors are self-limiting up to some reactivity insertion limit beyond which fuel damage occurs. This characteristic was studied in the Borax and Spert reactor tests of the 1950s and 1960s in the USA. This thesis considers the use of this experimental data in generic MTR-type reactor safety analysis. The approach presented herein is based on fundamental phenomenological understanding and uses correlations in the reactor test data with suitable account taken for differences in important system parameters. Specifically, a semi-empirical approach is used to quantify the relationship between the power, energy and temperature rise response of the system as well as parametric dependencies on void coefficient and the degree of subcooling. Secondary effects including the dependence on coolant flow are also examined. A rigorous curve fitting approach and error assessment is used to quantify the trends in the experimental data. In addition to the initial power burst stage of an unprotected transient, the longer term stability of the system is considered with a stylized treatment of characteristic power/temperature oscillations (chugging). A bridge from the HEU-based experimental data to the LEU fuel cycle is assessed and outlined based on existing simulation results presented in the literature. A cell-model based parametric study is included. The results are used to construct a practical safety analysis methodology for determining reactivity insertion safety limits for a light-water moderated and cooled MTR-type core.

  19. Nuclear reactor safety device

    DOEpatents

    Hutter, E.

    1983-08-15

    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.

  20. Evaluation of nuclear facility decommissioning projects. Summary report: North Carolina State University Research and Training Reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Link, B.W.; Miller, R.L.

    1983-08-01

    This document summarizes information from the decommissioning of the NCSUR-3 (R-3), a 10 KWt university research and training reactor. The decommissioning data were placed in a computerized information retrieval/manipulation system which permits future utilization of this information in pre-decommissioning activities with other university reactors of similar design. The information is presented both in some detail in its computer output form and also as a manually assembled summarization which highlights the more significant aspects of the decommissioning project. Decommissioning data from a generic study, NUREG/CR 1756, Technology, Safety and Costs of Decommissioning Nuclear Research and Test Reactors, and the decommissioning ofmore » the Ames Laboratory Research Reactor (ALRR), a 5 MWt research reactor, is also included for comparison.« less

  1. Safety approach to the selection of design criteria for the CRBRP reactor refueling system

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Meisl, C J; Berg, G E; Sharkey, N F

    1979-01-01

    The selection of safety design criteria for Liquid Metal Fast Breeder Reactor (LMFBR) refueling systems required the extrapolation of regulations and guidelines intended for Light Water Reactor refueling systems and was encumbered by the lack of benefit from a commercially licensed predecessor other than Fermi. The overall approach and underlying logic are described for developing safety design criteria for the reactor refueling system (RRS) of the Clinch River Breeder Reactor Plant (CRBRP). The complete selection process used to establish the criteria is presented, from the definition of safety functions to the finalization of safety design criteria in the appropriate documents.more » The process steps are illustrated by examples.« less

  2. Fundamental approaches for analysis thermal hydraulic parameter for Puspati Research Reactor

    NASA Astrophysics Data System (ADS)

    Hashim, Zaredah; Lanyau, Tonny Anak; Farid, Mohamad Fairus Abdul; Kassim, Mohammad Suhaimi; Azhar, Noraishah Syahirah

    2016-01-01

    The 1-MW PUSPATI Research Reactor (RTP) is the one and only nuclear pool type research reactor developed by General Atomic (GA) in Malaysia. It was installed at Malaysian Nuclear Agency and has reached the first criticality on 8 June 1982. Based on the initial core which comprised of 80 standard TRIGA fuel elements, the very fundamental thermal hydraulic model was investigated during steady state operation using the PARET-code. The main objective of this paper is to determine the variation of temperature profiles and Departure of Nucleate Boiling Ratio (DNBR) of RTP at full power operation. The second objective is to confirm that the values obtained from PARET-code are in agreement with Safety Analysis Report (SAR) for RTP. The code was employed for the hot and average channels in the core in order to calculate of fuel's center and surface, cladding, coolant temperatures as well as DNBR's values. In this study, it was found that the results obtained from the PARET-code showed that the thermal hydraulic parameters related to safety for initial core which was cooled by natural convection was in agreement with the designed values and safety limit in SAR.

  3. EU Funded Research Activities on NPPS Operational Safety

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Manolatos, P.; Van Goethem, G.

    2002-07-01

    The 5. framework programme (FP-5), the pluri-annual research programme of the European Union (EU), covers the period 1998-2002. Research on nuclear energy, fusion and fission, is covered by the EURATOM part of the FP-5. An overview of the Euratom's research on Nuclear Reactor Safety, managed by the DG-RTD of the European Commission (EC), is presented. This concerns 70 multi-partner projects of approximately euro 82.5 million total contract value that have been selected and co-financed during the period 1999-2001. They form the three clusters of projects dealing with the 'Operational Safety of Existing Installations'. 'Plant Life Extension and Management' (PLEM), 'Severemore » Accident Management' (SAM) and 'Evolutionary concepts' (EVOL). Emphasis is given here to the projects of the PLEM cluster. (authors)« less

  4. Preliminary study on new configuration with LEU fuel assemblies for the Dalat nuclear research reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Van Lam Pham; Vinh Vinh Le; Ton Nghiem Huynh

    2008-07-15

    The fuel conversion of the Dalat Nuclear Research Reactor (DNRR) is being realized. The DNRR is a pool type research reactor which was reconstructed from the 250 kW TRIGA- MARK II reactor. The reconstructed reactor attained its nominal power of 500 kW in February 1984. According to the results of design and safety analyses performed by the joint study between RERTR Program at Argonne National Laboratory (ANL) and Vietnam Atomic Energy Commission (VAEC) the mixed core of irradiated HEU and new LEU WWR-M2 fuel assemblies will be created soon. This paper presents the results of preliminary study on new configurationmore » with only LEU fuel assemblies for the DNRR. The codes MCNP, REBUS and VARI3D are used to calculate neutron flux performance in irradiation positions and kinetics parameters. The idea of change of Beryllium rod reloading enables to get working configuration assured shutdown margin, thermal-hydraulic safety and increase in thermal neutron flux in neutron trap at the center of DNRR active core. (author)« less

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

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

    2011-07-18

    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. Manymore » 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

  6. Gaseous fuel nuclear reactor research

    NASA Technical Reports Server (NTRS)

    Schwenk, F. C.; Thom, K.

    1975-01-01

    Gaseous-fuel nuclear reactors are described; their distinguishing feature is the use of fissile fuels in a gaseous or plasma state, thereby breaking the barrier of temperature imposed by solid-fuel elements. This property creates a reactor heat source that may be able to heat the propellant of a rocket engine to 10,000 or 20,000 K. At this temperature level, gas-core reactors would provide the breakthrough in propulsion needed to open the entire solar system to manned and unmanned spacecraft. The possibility of fuel recycling makes possible efficiencies of up to 65% and nuclear safety at reduced cost, as well as high-thrust propulsion capabilities with specific impulse up to 5000 sec.

  7. Study Gives Good Odds on Nuclear Reactor Safety

    ERIC Educational Resources Information Center

    Russell, Cristine

    1974-01-01

    Summarized is data from a recent study on nuclear reactor safety completed by Norman C. Rasmussen and others. Non-nuclear events are about 10,000 times more likely to produce large accidents than nuclear plants. (RH)

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

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

    1991-09-01

    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 reactormore » 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.« less

  9. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

  10. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

  11. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

  12. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

  13. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

  14. Strengthening IAEA Safeguards for Research Reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Reid, Bruce D.; Anzelon, George A.; Budlong-Sylvester, Kory

    During their December 10-11, 2013, workshop in Grenoble France, which focused on the history and future of safeguarding research reactors, the United States, France and the United Kingdom (UK) agreed to conduct a joint study exploring ways to strengthen the IAEA’s safeguards approach for declared research reactors. This decision was prompted by concerns about: 1) historical cases of non-compliance involving misuse (including the use of non-nuclear materials for production of neutron generators for weapons) and diversion that were discovered, in many cases, long after the violations took place and as part of broader pattern of undeclared activities in half amore » dozen countries; 2) the fact that, under the Safeguards Criteria, the IAEA inspects some reactors (e.g., those with power levels under 25 MWt) less than once per year; 3) the long-standing precedent of States using heavy water research reactors (HWRR) to produce plutonium for weapons programs; 4) the use of HEU fuel in some research reactors; and 5) various technical characteristics common to some types of research reactors that could provide an opportunity for potential proliferators to misuse the facility or divert material with low probability of detection by the IAEA. In some research reactors it is difficult to detect diversion or undeclared irradiation. In addition, infrastructure associated with research reactors could pose a safeguards challenge. To strengthen the effectiveness of safeguards at the State level, this paper advocates that the IAEA consider ways to focus additional attention and broaden its safeguards toolbox for research reactors. This increase in focus on the research reactors could begin with the recognition that the research reactor (of any size) could be a common path element on a large number of technically plausible pathways that must be considered when performing acquisition pathway analysis (APA) for developing a State Level Approach (SLA) and Annual Implementation Plan

  15. Flooding Experiments and Modeling for Improved Reactor Safety

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Solmos, M.; Hogan, K. J.; Vierow, K.

    2008-09-14

    Countercurrent two-phase flow and “flooding” phenomena in light water reactor systems are being investigated experimentally and analytically to improve reactor safety of current and future reactors. The aspects that will be better clarified are the effects of condensation and tube inclination on flooding in large diameter tubes. The current project aims to improve the level of understanding of flooding mechanisms and to develop an analysis model for more accurate evaluations of flooding in the pressurizer surge line of a Pressurized Water Reactor (PWR). Interest in flooding has recently increased because Countercurrent Flow Limitation (CCFL) in the AP600 pressurizer surge linemore » can affect the vessel refill rate following a small break LOCA and because analysis of hypothetical severe accidents with the current flooding models in reactor safety codes shows that these models represent the largest uncertainty in analysis of steam generator tube creep rupture. During a hypothetical station blackout without auxiliary feedwater recovery, should the hot leg become voided, the pressurizer liquid will drain to the hot leg and flooding may occur in the surge line. The flooding model heavily influences the pressurizer emptying rate and the potential for surge line structural failure due to overheating and creep rupture. The air-water test results in vertical tubes are presented in this paper along with a semi-empirical correlation for the onset of flooding. The unique aspects of the study include careful experimentation on large-diameter tubes and an integrated program in which air-water testing provides benchmark knowledge and visualization data from which to conduct steam-water testing.« less

  16. Research Program of a Super Fast Reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie

    2006-07-01

    Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is notmore » breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)« less

  17. Monte Carlo modelling of TRIGA research reactor

    NASA Astrophysics Data System (ADS)

    El Bakkari, B.; Nacir, B.; El Bardouni, T.; El Younoussi, C.; Merroun, O.; Htet, A.; Boulaich, Y.; Zoubair, M.; Boukhal, H.; Chakir, M.

    2010-10-01

    The Moroccan 2 MW TRIGA MARK II research reactor at Centre des Etudes Nucléaires de la Maâmora (CENM) achieved initial criticality on May 2, 2007. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes for their use in agriculture, industry, and medicine. This study deals with the neutronic analysis of the 2-MW TRIGA MARK II research reactor at CENM and validation of the results by comparisons with the experimental, operational, and available final safety analysis report (FSAR) values. The study was prepared in collaboration between the Laboratory of Radiation and Nuclear Systems (ERSN-LMR) from Faculty of Sciences of Tetuan (Morocco) and CENM. The 3-D continuous energy Monte Carlo code MCNP (version 5) was used to develop a versatile and accurate full model of the TRIGA core. The model represents in detailed all components of the core with literally no physical approximation. Continuous energy cross-section data from the more recent nuclear data evaluations (ENDF/B-VI.8, ENDF/B-VII.0, JEFF-3.1, and JENDL-3.3) as well as S( α, β) thermal neutron scattering functions distributed with the MCNP code were used. The cross-section libraries were generated by using the NJOY99 system updated to its more recent patch file "up259". The consistency and accuracy of both the Monte Carlo simulation and neutron transport physics were established by benchmarking the TRIGA experiments. Core excess reactivity, total and integral control rods worth as well as power peaking factors were used in the validation process. Results of calculations are analysed and discussed.

  18. Tower Shielding Reactor II design and operation report: Vol. 2. Safety Analysis

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Holland, L. B.; Kolb, J. O.

    1970-01-01

    Information on the Tower Shielding Reactor II is contained in the TSR-II Design and Operation Report and in the Tower Shielding Facility Manual. The TSR-II Design and Operating Report consists of three volumes. Volume 1 is Descriptions of the Tower Shielding Reactor II and Facility; Volume 2 is Safety analysis of the Tower Shielding Reactor II; and Volume 3 is the Assembly and Testing of the Tower Shielding Reactor II Control Mechanism Housing.

  19. Experimental study of radiation dose rate at different strategic points of the BAEC TRIGA Research Reactor.

    PubMed

    Ajijul Hoq, M; Malek Soner, M A; Salam, M A; Haque, M M; Khanom, Salma; Fahad, S M

    2017-12-01

    The 3MW TRIGA Mark-II Research Reactor of Bangladesh Atomic Energy Commission (BAEC) has been under operation for about thirty years since its commissioning at 1986. In accordance with the demand of fundamental nuclear research works, the reactor has to operate at different power levels by utilizing a number of experimental facilities. Regarding the enquiry for safety of reactor operating personnel and radiation workers, it is necessary to know the radiation level at different strategic points of the reactor where they are often worked. In the present study, neutron, beta and gamma radiation dose rate at different strategic points of the reactor facility with reactor power level of 2.4MW was measured to estimate the rising level of radiation due to its operational activities. From the obtained results high radiation dose is observed at the measurement position of the piercing beam port which is caused by neutron leakage and accordingly, dose rate at the stated position with different reactor power levels was measured. This study also deals with the gamma dose rate measurements at a fixed position of the reactor pool top surface for different reactor power levels under both Natural Convection Cooling Mode (NCCM) and Forced Convection Cooling Mode (FCCM). Results show that, radiation dose rate is higher for NCCM in compared with FCCM and increasing with the increase of reactor power. Thus, concerning the radiological safety issues for working personnel and the general public, the radiation dose level monitoring and the experimental analysis performed within this paper is so much effective and the result of this work can be utilized for base line data and code verification of the nuclear reactor. Copyright © 2017 Elsevier Ltd. All rights reserved.

  20. Reliability and safety of the electrical power supply complex of the Hanford production reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Robbins, F.D.

    Safety has been and must continue to be the inviolable modulus by which the operation of a nuclear reactor must be judged. A malfunction in any reactor may well result in a release of fission products which may dissipate over a wide geographical area. Such dissipation may place the health, happiness and even the lives of the people in the region in serious jeopardy. As a result, the property damage and liability cost may reach astronomical values in the order of magnitude of billions of dollars. Reliability of the electrical network is an indispensable factor in attaining a high ordermore » of safety assurance. Progress in the peaceful use of atomic energy may take the form of electrical power generation using the nuclear reactor as a source of thermal energy. In view of these factors it seems appropriate and profitable that a critical engineering study be made of the safety and reliability of the Hanford reactors without regard to cost economics. This individual and independent technical engineering analysis was made without regard to Hanford traditional engineering and administration assignments. The main objective has been to focus attention on areas which seem to merit further detailed study on conditions which seem to need adjustment but most of all on those changes which will improve reactor safety. This report is the result of such a study.« less

  1. Analysis of Loss-of-Coolant Accidents in the NIST Research Reactor - Early Phase

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Baek, Joo S.; Diamond, David

    A study of the fuel temperature during the early phase of a loss-of-coolant accident (LOCA) in the NIST research reactor (NBSR) was completed. Previous studies had been reported in the preliminary safety analysis report for the conversion of the NBSR from high-enriched uranium (HEU) fuel to low-enriched (LEU) fuel. Those studies had focused on the most vulnerable LOCA situation, namely, a double-ended guillotine break in the time period after reactor trip when water is drained from either the coolant channels inside the fuel elements or the region outside the fuel elements. The current study fills in a gap in themore » analysis which is the early phase of the event when there may still be water present but the reactor is at power or immediately after reactor trip and pumps have tripped. The calculations were done, for both the current HEU-fueled core and the proposed LEU core, with the TRACE thermal-hydraulic systems code. Several break locations and different break sizes were considered. In all cases the increase in the clad (or fuel meat) temperature was relatively small so that a large margin to the temperature threshold for blistering (the Safety Limit for the NBSR) remained.« less

  2. Supply of enriched uranium for research reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Mueller, H.

    1997-08-01

    Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel onmore » December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA`s ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future.« less

  3. 2007 international meeting on Reduced Enrichment for Research and Test Reactors (RERTR). Abstracts and available papers presented at the meeting

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    NONE

    2008-07-15

    The Meeting papers discuss research and test reactor fuel performance, manufacturing and testing. Some of the main topics are: conversion from HEU to LEU in different reactors and corresponding problems and activities; flux performance and core lifetime analysis with HEU and LEU fuels; physics and safety characteristics; measurement of gamma field parameters in core with LEU fuel; nondestructive analysis of RERTR fuel; thermal hydraulic analysis; fuel interactions; transient analyses and thermal hydraulics for HEU and LEU cores; microstructure research reactor fuels; post irradiation analysis and performance; computer codes and other related problems.

  4. Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Mourao, Rogerio P.; Leite da Silva, Luiz; Miranda, Carlos A.

    2013-07-01

    Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away frommore » reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)« less

  5. Revitalizing Nuclear Safety Research.

    ERIC Educational Resources Information Center

    National Academy of Sciences - National Research Council, Washington, DC.

    This report covers the general issues involved in nuclear safety research and points out the areas needing detailed consideration. Topics included are: (1) "Principles of Nuclear Safety Research" (examining who should fund, who should conduct, and who should set the agenda for nuclear safety research); (2) "Elements of a Future…

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Smith, M. C.; Wright, R. F.

    2012-07-01

    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. Themore » 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

  7. IAEA Coordinated Research Project on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Strydom, Gerhard; Bostelmann, F.

    The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal-hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis (SA) and uncertainty analysis (UA) methods. Uncertainty originates from errors in physical data, manufacturing uncertainties, modelling and computational algorithms. (The interested reader is referred to the large body of published SA and UA literature for a more complete overview of the various types of uncertainties, methodologies and results obtained).more » SA is helpful for ranking the various sources of uncertainty and error in the results of core analyses. SA and UA are required to address cost, safety, and licensing needs and should be applied to all aspects of reactor multi-physics simulation. SA and UA can guide experimental, modelling, and algorithm research and development. Current SA and UA rely either on derivative-based methods such as stochastic sampling methods or on generalized perturbation theory to obtain sensitivity coefficients. Neither approach addresses all needs. In order to benefit from recent advances in modelling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Only a parallel effort in advanced simulation and in nuclear data improvement will be able to provide designers with more robust and well validated calculation tools to meet design target accuracies. In February 2009, the Technical Working Group on Gas-Cooled Reactors (TWG-GCR) of the International Atomic Energy Agency (IAEA) recommended that the proposed Coordinated Research Program

  8. Experiences in utilization of research reactors in Yugoslavia

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Copic, M.; Gabrovsek, Z.; Pop-Jordanov, J.

    1971-06-15

    The nuclear institutes in Yugoslavia possess three research reactors. Since 1958, two heavy-water reactors have been in operation at the 'Boris Kidric' Institute, a zero-power reactor RB and a 6. 5-MW reactor RA. At the Jozef Stefan Institute, a 250-kW TRIGA Mark II reactor has been operating since 1966. All reactors are equipped with the necessary experimental facilities. The main activities based on these reactors are: (1) fundamental research in solid-state and nuclear physics; (2) R and D activities related to nuclear power program; and (3) radioisotope production. In fundamental physics, inelastic neutron scattering and diffraction phenomena are studied bymore » means of the neutron beam tubes and applied to investigations of the structures of solids and liquids. Valuable results are also obtained in n - γ reaction studies. Experiments connected with the fuel -element development program, owing to the characteristics of the existing reactors, are limited to determination of the fuel element parameters, to studies on the purity of uranium, and to a small number of capsule irradiations. All three reactors are also used for the verification of different methods applied in the analysis of power reactors, particularly concerning neutron flux distributions, the optimization of reactor core configurations and the shielding effects. An appreciable irradiation space in the reactors is reserved for isotope production. Fruitful international co-operation has been established in all these activities, on the basis of either bilateral or multilateral arrangements. The paper gives a critical analysis of the utilization of research reactors in a developing country such as Yugoslavia. The investments in and the operational costs of research reactors are compared with the benefits obtained in different areas of reactor application. The impact on the general scientific, technological and educational level in the country is also considered. In particular, an attempt is made ro

  9. Status report on the Small Secure Transportable Autonomous Reactor (SSTAR) /Lead-cooled Fast Reactor (LFR) and supporting research and development.

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Sienicki, J. J.; Moisseytsev, A.; Yang, W. S.

    2008-06-23

    This report provides an update on development of a pre-conceptual design for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) plant concept and supporting research and development activities. SSTAR is a small, 20 MWe (45 MWt), natural circulation, fast reactor plant for international deployment concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety further supporting deployment in developing nations. In FY 2006, improvements have been mademore » at ANL to the pre-conceptual design of both the reactor system and the energy converter which incorporates a supercritical carbon dioxide Brayton cycle providing higher plant efficiency (44 %) and improved economic competitiveness. The supercritical CO2 Brayton cycle technology is also applicable to Sodium-Cooled Fast Reactors providing the same benefits. One key accomplishment has been the development of a control strategy for automatic control of the supercritical CO2 Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power. Under autonomous load following operation, the reactor core power adjusts itself to equal the heat removal from the reactor system to the power converter through the large reactivity feedback of the fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the reactor to the grid load. The report includes early calculations for an international benchmarking problem for a LBE-cooled, nitride-fueled fast reactor core organized by the IAEA as part of a Coordinated Research Project on Small Reactors without Onsite Refueling; the calculations use the same

  10. Manned space flight nuclear system safety. Volume 3: Reactor system preliminary nuclear safety analysis. Part 1: Reference Design Document (RDD)

    NASA Technical Reports Server (NTRS)

    1972-01-01

    The Reference Design Document, of the Preliminary Safety Analysis Report (PSAR) - Reactor System provides the basic design and operations data used in the nuclear safety analysis of the Rector Power Module as applied to a Space Base program. A description of the power module systems, facilities, launch vehicle and mission operations, as defined in NASA Phase A Space Base studies is included. Each of two Zirconium Hydride Reactor Brayton power modules provides 50 kWe for the nominal 50 man Space Base. The INT-21 is the prime launch vehicle. Resupply to the 500 km orbit over the ten year mission is provided by the Space Shuttle. At the end of the power module lifetime (nominally five years), a reactor disposal system is deployed for boost into a 990 km high altitude (long decay time) earth orbit.

  11. Safety Issues at the Defense Production Reactors. A Report to the U.S. Department of Energy.

    ERIC Educational Resources Information Center

    National Academy of Sciences - National Research Council, Washington, DC. Commission on Physical Sciences, Mathematics, and Resources.

    This report provides an assessment of safety management, safety review, and safety methodology employed by the Department of Energy (DOE) and private contractors. Chapter 1, "The DOE Safety Framework," examines safety objectives for production reactors and processes to implement the objectives. Chapter 2, "Technical Issues,"…

  12. 78 FR 58575 - Review of Experiments for Research Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-09-24

    ... NUCLEAR REGULATORY COMMISSION [NRC-2013-0219] Review of Experiments for Research Reactors AGENCY... Commission (NRC) is withdrawing Regulatory Guide (RG) 2.4, ``Review of Experiments for Research Reactors... withdrawing RG 2.4, ``Review of Experiments for Research Reactors,'' (ADAMS Accession No. ML003740131) because...

  13. A reload and startup plan for conversion of the NIST research reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    D. J. Diamond

    The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts. The reloadmore » portion of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.« less

  14. 77 FR 60479 - Burnup Credit in the Criticality Safety Analyses of Pressurized Water Reactor Spent Fuel in...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-03

    ... Pressurized Water Reactor Spent Fuel in Transportation and Storage Casks AGENCY: Nuclear Regulatory Commission... 3, entitled, ``Burnup Credit in the Criticality Safety Analyses of PWR [Pressurized Water Reactor... water reactor spent nuclear fuel (SNF) in transportation packages and storage casks. SFST-ISG-8...

  15. GROWTH OF THE INTERNATIONAL CRITICALITY SAFETY AND REACTOR PHYSICS EXPERIMENT EVALUATION PROJECTS

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    J. Blair Briggs; John D. Bess; Jim Gulliford

    2011-09-01

    Since the International Conference on Nuclear Criticality Safety (ICNC) 2007, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) have continued to expand their efforts and broaden their scope. Eighteen countries participated on the ICSBEP in 2007. Now, there are 20, with recent contributions from Sweden and Argentina. The IRPhEP has also expanded from eight contributing countries in 2007 to 16 in 2011. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments1' have increased from 442 evaluations (38000 pages), containing benchmark specifications for 3955 critical ormore » subcritical configurations to 516 evaluations (nearly 55000 pages), containing benchmark specifications for 4405 critical or subcritical configurations in the 2010 Edition of the ICSBEP Handbook. The contents of the Handbook have also increased from 21 to 24 criticality-alarm-placement/shielding configurations with multiple dose points for each, and from 20 to 200 configurations categorized as fundamental physics measurements relevant to criticality safety applications. Approximately 25 new evaluations and 150 additional configurations are expected to be added to the 2011 edition of the Handbook. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments2' have increased from 16 different experimental series that were performed at 12 different reactor facilities to 53 experimental series that were performed at 30 different reactor facilities in the 2011 edition of the Handbook. Considerable effort has also been made to improve the functionality of the searchable database, DICE (Database for the International Criticality Benchmark Evaluation Project) and verify the accuracy of the data contained therein. DICE will be discussed in separate papers at ICNC 2011. The status of the ICSBEP and

  16. Analysis of radiation safety for Small Modular Reactor (SMR) on PWR-100 MWe type

    NASA Astrophysics Data System (ADS)

    Udiyani, P. M.; Husnayani, I.; Deswandri; Sunaryo, G. R.

    2018-02-01

    Indonesia as an archipelago country, including big, medium and small islands is suitable to construction of Small Medium/Modular reactors. Preliminary technology assessment on various SMR has been started, indeed the SMR is grouped into Light Water Reactor, Gas Cooled Reactor, and Solid Cooled Reactor and from its site it is group into Land Based reactor and Water Based Reactor. Fukushima accident made people doubt about the safety of Nuclear Power Plant (NPP), which impact on the public perception of the safety of nuclear power plants. The paper will describe the assessment of safety and radiation consequences on site for normal operation and Design Basis Accident postulation of SMR based on PWR-100 MWe in Bangka Island. Consequences of radiation for normal operation simulated for 3 units SMR. The source term was generated from an inventory by using ORIGEN-2 software and the consequence of routine calculated by PC-Cream and accident by PC Cosyma. The adopted methodology used was based on site-specific meteorological and spatial data. According to calculation by PC-CREAM 08 computer code, the highest individual dose in site area for adults is 5.34E-02 mSv/y in ESE direction within 1 km distance from stack. The result of calculation is that doses on public for normal operation below 1mSv/y. The calculation result from PC Cosyma, the highest individual dose is 1.92.E+00 mSv in ESE direction within 1km distance from stack. The total collective dose (all pathway) is 3.39E-01 manSv, with dominant supporting from cloud pathway. Results show that there are no evacuation countermeasure will be taken based on the regulation of emergency.

  17. The Simulator Development for RDE Reactor

    NASA Astrophysics Data System (ADS)

    Subekti, Muhammad; Bakhri, Syaiful; Sunaryo, Geni Rina

    2018-02-01

    BATAN is proposing the construction of experimental power reactor (RDE reactor) for increasing the public acceptance on NPP development plan, proofing the safety level of the most advanced reactor by performing safety demonstration on the accidents such as Chernobyl and Fukushima, and owning the generation fourth (G4) reactor technology. For owning the reactor technology, the one of research activities is RDE’s simulator development that employing standard equation. The development utilizes standard point kinetic and thermal equation. The examination of the simulator carried out comparison in which the simulation’s calculation result has good agreement with assumed parameters and ChemCAD calculation results. The transient simulation describes the characteristic of the simulator to respond the variation of power increase of 1.5%/min, 2.5%/min, and 3.5%/min.

  18. Safety and licensing of a small modular gas-cooled reactor system

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Brown, N.W.; Kelley, A.P. Jr.

    A modular side-by-side high-temperature gas-cooled reactor (SBS-HTGR) is being developed by Interatom/Kraftwerk Union (KWU). The General Electric Company and Interatom/KWU entered into a proprietary working agreement to continue develop jointly of the SBS-HTGR. A study on adapting the SBS-HTGR for application in the US has been completed. The study investigated the safety characteristics and the use of this type of design in an innovative approach to licensing. The safety objective guiding the design of the modular SBS-HTGR is to control radionuclide release by the retention of fission products within the fuel particles with minimal reliance on active design features. Themore » philosophy on which this objective is predicated is that by providing a simple safety case, the safety criteria can be demonstrated as being met with high confidence through conduct of a full-scale module safety test.« less

  19. RELAP5 Application to Accident Analysis of the NIST Research Reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Baek, J.; Cuadra Gascon, A.; Cheng, L.Y.

    Detailed safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The time-dependent analysis of the primary system is determined with a RELAP5 transient analysis model that includes the reactor vessel, the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. A post-processing of the simulation results has been conducted to evaluate minimum critical heat flux ratio (CHFR) using the Sudo-Kaminaga correlation. Evaluations are performed for the following accidents: (1) the control rod withdrawal startup accidentmore » and (2) the maximum reactivity insertion accident. In both cases the RELAP5 results indicate that there is adequate margin to CHF and no damage to the fuel will occur because of sufficient coolant flow through the fuel channels and the negative scram reactivity insertion.« less

  20. Hydrogen and water reactor safety: proceedings

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Not Available

    1982-01-01

    Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability.

  1. Advanced Fuel Cycles for Fusion Reactors: Passive Safety and Zero-Waste Options

    NASA Astrophysics Data System (ADS)

    Zucchetti, Massimo; Sugiyama, Linda E.

    2006-05-01

    Nuclear fusion is seen as a much ''cleaner'' energy source than fission. Most of the studies and experiments on nuclear fusion are currently devoted to the Deuterium-Tritium (DT) fuel cycle, since it is the easiest way to reach ignition. The recent stress on safety by the world's community has stimulated the research on other fuel cycles than the DT one, based on 'advanced' reactions, such as the Deuterium-Helium-3 (DHe) one. These reactions pose problems, such as the availability of 3He and the attainment of the higher plasma parameters that are required for burning. However, they have many advantages, like for instance the very low neutron activation, while it is unnecessary to breed and fuel tritium. The extrapolation of Ignitor technologies towards a larger and more powerful experiment using advanced fuel cycles (Candor) has been studied. Results show that Candor does reach the passive safety and zero-waste option. A fusion power reactor based on the DHe cycle could be the ultimate response to the environmental requirements for future nuclear power plants.

  2. An overview of research activities on materials for nuclear applications at the INL Safety, Tritium and Applied Research facility

    NASA Astrophysics Data System (ADS)

    Calderoni, P.; Sharpe, J.; Shimada, M.; Denny, B.; Pawelko, B.; Schuetz, S.; Longhurst, G.; Hatano, Y.; Hara, M.; Oya, Y.; Otsuka, T.; Katayama, K.; Konishi, S.; Noborio, K.; Yamamoto, Y.

    2011-10-01

    The Safety, Tritium and Applied Research facility at the Idaho National Laboratory is a US Department of Energy National User Facility engaged in various aspects of materials research for nuclear applications related to fusion and advanced fission systems. Research activities are mainly focused on the interaction of tritium with materials, in particular plasma facing components, liquid breeders, high temperature coolants, fuel cladding, cooling and blanket structures and heat exchangers. Other activities include validation and verification experiments in support of the Fusion Safety Program, such as beryllium dust reactivity and dust transport in vacuum vessels, and support of Advanced Test Reactor irradiation experiments. This paper presents an overview of the programs engaged in the activities, which include the US-Japan TITAN collaboration, the US ITER program, the Next Generation Power Plant program and the tritium production program, and a presentation of ongoing experiments as well as a summary of recent results with emphasis on fusion relevant materials.

  3. Researchers' Roles in Patient Safety Improvement.

    PubMed

    Pietikäinen, Elina; Reiman, Teemu; Heikkilä, Jouko; Macchi, Luigi

    2016-03-01

    In this article, we explore how researchers can contribute to patient safety improvement. We aim to expand the instrumental role researchers have often occupied in relation to patient safety improvement. We reflect on our own improvement model and experiences as patient safety researchers in an ongoing Finnish multi-actor innovation project through self-reflective narration. Our own patient safety improvement model can be described as systemic. Based on the purpose of the innovation project, our improvement model, and the improvement models of the other actors in the project, we have carried out a wide range of activities. Our activities can be summarized in 8 overlapping patient safety improvement roles: modeler, influencer, supplier, producer, ideator, reflector, facilitator, and negotiator. When working side by side with "practice," researchers are offered and engage in several different activities. The way researchers contribute to patient safety improvement and balance between different roles depends on the purpose of the study, as well as on the underlying patient safety improvement models. Different patient safety research paradigms seem to emphasize different improvement roles, and thus, they also face different challenges. Open reflection on the underlying improvement models and roles can help researchers with different backgrounds-as well as other actors involved in patient safety improvement-in structuring their work and collaborating productively.

  4. SLSF in-reactor local fault safety experiment P4. Final report

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Thompson, D. H.; Holland, J. W.; Braid, T. H.

    The Sodium Loop Safety Facility (SLSF), a major facility in the US fast-reactor safety program, has been used to simulate a variety of sodium-cooled fast reactor accidents. SLSF experiment P4 was conducted to investigate the behavior of a "worse-than-case" local fault configuration. Objectives of this experiment were to eject molten fuel into a 37-pin bundle of full-length Fast-Test-Reactor-type fuel pins form heat-generating fuel canisters, to characterize the severity of any molten fuel-coolant interaction, and to demonstrate that any resulting blockage could either be tolerated during continued power operation or detected by global monitors to prevent fuel failure propagation. The designmore » goal for molten fuel release was 10 to 30 g. Explusion of molten fuel from fuel canisters caused failure of adjacent pins and a partial flow channel blockage in the fuel bundle during full-power operation. Molten fuel and fuel debris also lodged against the inner surface of the test subassembly hex-can wall. The total fuel disruption of 310 g evaluated from posttest examination data was in excellent agreement with results from the SLSF delayed neutron detection system, but exceeded the target molten fuel release by an order of magnitude. This report contains a summary description of the SLSF in-reactor loop and support systems and the experiment operations. results of the detailed macro- and microexamination of disrupted fuel and metal and results from the analysis of the on-line experimental data are described, as are the interpretations and conclusions drawn from the posttest evaluations. 60 refs., 74 figs.« less

  5. A Reload and Startup Plan for and #8233;Conversion of the NIST Research Reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Diamond, D. J.; Varuttamaseni, A.

    The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts.The reload portionmore » of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.« less

  6. A Potential NASA Research Reactor to Support NTR Development

    NASA Technical Reports Server (NTRS)

    Eades, Michael; Gerrish, Harold; Hardin, Leroy

    2013-01-01

    In support of efforts for research into the design and development of a man rated Nuclear Thermal Rocket (NTR) engine, the National Aeronautics and Space Administration (NASA), Marshall Space Flight Center (MSFC), is evaluating the potential for building a Nuclear Regulatory Commission (NRC) licensed research reactor. The proposed reactor would be licensed by NASA and operated jointly by NASA and university partners. The purpose of this reactor would be to perform further research into the technologies and systems needed for a successful NTR project and promote nuclear training and education.

  7. The Australian Replacement Research Reactor

    NASA Astrophysics Data System (ADS)

    Kennedy, Shane; Robinson, Robert

    2004-03-01

    The 20-MW Australian Replacement Research Reactor represents possibly the greatest single research infrastructure investment in Australia's history. Construction of the facility has commenced, following award of the construction contract in July 2000, and the construction licence in April 2002. The project includes a large state-of-the-art liquid deuterium cold-neutron source and supermirror guides feeding a large modern guide hall, in which most of the instruments are placed. Alongside the guide hall, there is good provision of laboratory, office and space for support activities. While the facility has "space" for up to 18 instruments, the project has funding for an initial set of 8 instruments, which will be ready when the reactor is fully operational in July 2006. Instrument performance will be competitive with the best research-reactor facilities anywhere, and our goal is to be in the top 3 such facilities worldwide. Staff to lead the design effort and man these instruments have been hired on the international market from leading overseas facilities, and from within Australia, and 7 out of 8 instruments have been specified and costed. At present the instrumentation project carries 10contingency. An extensive dialogue has taken place with the domestic user community and our international peers, via various means including a series of workshops over the last 2 years covering all 8 instruments, emerging areas of application like biology and the earth sciences, and computing infrastructure for the instruments.

  8. Advanced Concepts for Pressure-Channel Reactors: Modularity, Performance and Safety

    NASA Astrophysics Data System (ADS)

    Duffey, Romney B.; Pioro, Igor L.; Kuran, Sermet

    Based on an analysis of the development of advanced concepts for pressure-tube reactor technology, we adapt and adopt the pressure-tube reactor advantage of modularity, so that the subdivided core has the potential for optimization of the core, safety, fuel cycle and thermal performance independently, while retaining passive safety features. In addition, by adopting supercritical water-cooling, the logical developments from existing supercritical turbine technology and “steam” systems can be utilized. Supercritical and ultra-supercritical boilers and turbines have been operating for some time in coal-fired power plants. Using coolant outlet temperatures of about 625°C achieves operating plant thermal efficiencies in the order of 45-48%, using a direct turbine cycle. In addition, by using reheat channels, the plant has the potential to produce low-cost process heat, in amounts that are customer and market dependent. The use of reheat systems further increases the overall thermal efficiency to 55% and beyond. With the flexibility of a range of plant sizes suitable for both small (400 MWe) and large (1400 MWe) electric grids, and the ability for co-generation of electric power, process heat, and hydrogen, the concept is competitive. The choice of core power, reheat channel number and exit temperature are all set by customer and materials requirements. The pressure channel is a key technology that is needed to make use of supercritical water (SCW) in CANDU®1 reactors feasible. By optimizing the fuel bundle and fuel channel, convection and conduction assure heat removal using passive-moderator cooling. Potential for severe core damage can be almost eliminated, even without the necessity of activating the emergency-cooling systems. The small size of containment structure lends itself to a small footprint, impacts economics and building techniques. Design features related to Canadian concepts are discussed in this paper. The main conclusion is that development of

  9. Background radiation measurements at high power research reactors

    NASA Astrophysics Data System (ADS)

    Ashenfelter, J.; Balantekin, B.; Baldenegro, C. X.; Band, H. R.; Barclay, G.; Bass, C. D.; Berish, D.; Bowden, N. S.; Bryan, C. D.; Cherwinka, J. J.; Chu, R.; Classen, T.; Davee, D.; Dean, D.; Deichert, G.; Dolinski, M. J.; Dolph, J.; Dwyer, D. A.; Fan, S.; Gaison, J. K.; Galindo-Uribarri, A.; Gilje, K.; Glenn, A.; Green, M.; Han, K.; Hans, S.; Heeger, K. M.; Heffron, B.; Jaffe, D. E.; Kettell, S.; Langford, T. J.; Littlejohn, B. R.; Martinez, D.; McKeown, R. D.; Morrell, S.; Mueller, P. E.; Mumm, H. P.; Napolitano, J.; Norcini, D.; Pushin, D.; Romero, E.; Rosero, R.; Saldana, L.; Seilhan, B. S.; Sharma, R.; Stemen, N. T.; Surukuchi, P. T.; Thompson, S. J.; Varner, R. L.; Wang, W.; Watson, S. M.; White, B.; White, C.; Wilhelmi, J.; Williams, C.; Wise, T.; Yao, H.; Yeh, M.; Yen, Y.-R.; Zhang, C.; Zhang, X.; Prospect Collaboration

    2016-01-01

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including γ-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the background fields encountered. The general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.

  10. Research reactor decommissioning experience - concrete removal and disposal -

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Manning, Mark R.; Gardner, Frederick W.

    1990-07-01

    Removal and disposal of neutron activated concrete from biological shields is the most significant operational task associated with research reactor decommissioning. During the period of 1985 thru 1989 Chem-Nuclear Systems, Inc. was the prime contractor for complete dismantlement and decommissioning of the Northrop TRIGA Mark F, the Virginia Tech Argonaut, and the Michigan State University TRIGA Mark I Reactor Facilities. This paper discusses operational requirements, methods employed, and results of the concrete removal, packaging, transport and disposal operations for these (3) research reactor decommissioning projects. Methods employed for each are compared. Disposal of concrete above and below regulatory release limitsmore » for unrestricted use are discussed. This study concludes that activated reactor biological shield concrete can be safely removed and buried under current regulations.« less

  11. Safety and core design of large liquid-metal cooled fast breeder reactors

    NASA Astrophysics Data System (ADS)

    Qvist, Staffan Alexander

    In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

  12. Opportunities for Materials Science and Biological Research at the OPAL Research Reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Kennedy, S. J.

    Neutron scattering techniques have evolved over more than 1/2 century into a powerful set of tools for determination of atomic and molecular structures. Modern facilities offer the possibility to determine complex structures over length scales from {approx}0.1 nm to {approx}500 nm. They can also provide information on atomic and molecular dynamics, on magnetic interactions and on the location and behaviour of hydrogen in a variety of materials. The OPAL Research Reactor is a 20 megawatt pool type reactor using low enriched uranium fuel, and cooled by water. OPAL is a multipurpose neutron factory with modern facilities for neutron beam research,more » radioisotope production and irradiation services. The neutron beam facility has been designed to compete with the best beam facilities in the world. After six years in construction, the reactor and neutron beam facilities are now being commissioned, and we will commence scientific experiments later this year. The presentation will include an outline of the strengths of neutron scattering and a description of the OPAL research reactor, with particular emphasis on it's scientific infrastructure. It will also provide an overview of the opportunities for research in materials science and biology that will be possible at OPAL, and mechanisms for accessing the facilities. The discussion will emphasize how researchers from around the world can utilize these exciting new facilities.« less

  13. Background radiation measurements at high power research reactors

    DOE PAGES

    Ashenfelter, J.; Yeh, M.; Balantekin, B.; ...

    2015-10-23

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including γ-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the backgroundmore » fields encountered. Furthermore, the general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.« less

  14. Evaluating the Cost, Safety, and Proliferation Risks of Small Floating Nuclear Reactors.

    PubMed

    Ford, Michael J; Abdulla, Ahmed; Morgan, M Granger

    2017-11-01

    It is hard to see how our energy system can be decarbonized if the world abandons nuclear power, but equally hard to introduce the technology in nonnuclear energy states. This is especially true in countries with limited technical, institutional, and regulatory capabilities, where safety and proliferation concerns are acute. Given the need to achieve serious emissions mitigation by mid-century, and the multidecadal effort required to develop robust nuclear governance institutions, we must look to other models that might facilitate nuclear plant deployment while mitigating the technology's risks. One such deployment paradigm is the build-own-operate-return model. Because returning small land-based reactors containing spent fuel is infeasible, we evaluate the cost, safety, and proliferation risks of a system in which small modular reactors are manufactured in a factory, and then deployed to a customer nation on a floating platform. This floating small modular reactor would be owned and operated by a single entity and returned unopened to the developed state for refueling. We developed a decision model that allows for a comparison of floating and land-based alternatives considering key International Atomic Energy Agency plant-siting criteria. Abandoning onsite refueling is beneficial, and floating reactors built in a central facility can potentially reduce the risk of cost overruns and the consequences of accidents. However, if the floating platform must be built to military-grade specifications, then the cost would be much higher than a land-based system. The analysis tool presented is flexible, and can assist planners in determining the scope of risks and uncertainty associated with different deployment options. © 2017 Society for Risk Analysis.

  15. Dutch pedestrian safety research review

    DOT National Transportation Integrated Search

    1999-12-01

    This report is one in a series of pedestrian safety synthesis reports prepared for the Federal Highway Administration to document pedestrian safety in other countries. This report reviews recent pedestrian safety research in the Netherlands. It addre...

  16. Safety management of complex research operations

    NASA Technical Reports Server (NTRS)

    Brown, W. J.

    1981-01-01

    Complex research and technology operations present many varied potential hazards which must be addressed in a disciplined independent safety review and approval process. The research and technology effort at the Lewis Research Center is divided into programmatic areas of aeronautics, space and energy. Potential hazards vary from high energy fuels to hydrocarbon fuels, high pressure systems to high voltage systems, toxic chemicals to radioactive materials and high speed rotating machinery to high powered lasers. A Safety Permit System presently covers about 600 potentially hazardous operations. The Safety Management Program described in this paper is believed to be a major factor in maintaining an excellent safety record at the Lewis Research Center.

  17. Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Licht, J.; Bergeron, A.; Dionne, B.

    2016-01-01

    Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimentalmore » device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.« less

  18. Canadian research on pedestrian safety

    DOT National Transportation Integrated Search

    1999-12-01

    This report is one in a series of pedestrian safety synthesis reports prepared for the Federal Highway Administration to document pedestrian safety in other countries. This report reviews Canadian research in six areas of pedestrian safety: (1) Inter...

  19. A review of inherent safety characteristics of metal alloy sodium-cooled fast reactor fuel against postulated accidents

    DOE PAGES

    Sofu, Tanju

    2015-04-01

    The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, double-fault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperaturemore » profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain coolable. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel--coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.« less

  20. A review of inherent safety characteristics of metal alloy sodium-cooled fast reactor fuel against postulated accidents

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Sofu, Tanju

    2015-04-01

    The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, double-fault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperaturemore » profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain cool-able. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.« less

  1. BESAFE II: Accident safety analysis code for MFE reactor designs

    NASA Astrophysics Data System (ADS)

    Sevigny, Lawrence Michael

    The viability of controlled thermonuclear fusion as an alternative energy source hinges on its desirability from an economic and an environmental and safety standpoint. It is the latter which is the focus of this thesis. For magnetic fusion energy (MFE) devices, the safety concerns equate to a design's behavior during a worst-case accident scenario which is the loss of coolant accident (LOCA). In this dissertation, we examine the behavior of MFE devices during a LOCA and how this behavior relates to the safety characteristics of the machine; in particular the acute, whole-body, early dose. In doing so, we have produced an accident safety code, BESAFE II, now available to the fusion reactor design community. The Appendix constitutes the User's Manual for BESAFE II. The theory behind early dose calculations including the mobilization of activation products is presented in Chapter 2. Since mobilization of activation products is a strong function of temperature, it becomes necessary to calculate the thermal response of a design during a LOCA in order to determine the fraction of the activation products which are mobilized and thus become the source for the dose. The code BESAFE II is designed to determine the temperature history of each region of a design and determine the resulting mobilization of activation products at each point in time during the LOCA. The BESAFE II methodology is discussed in Chapter 4, followed by demonstrations of its use for two reference design cases: a PCA-Li tokamak and a SiC-He tokamak. Of these two cases, it is shown that the SiC-He tokamak is a better design from an accident safety standpoint than the PCA-Li tokamak. It is also found that doses derived from temperature-dependent mobilization data are different than those predicted using set mobilization categories such as those that involve Piet fractions. This demonstrates the need for more experimental data on fusion materials. The possibility for future improvements and modifications

  2. Safety apparatus for nuclear reactor to prevent structural damage from overheating by core debris

    DOEpatents

    Gabor, John D.; Cassulo, John C.; Pedersen, Dean R.; Baker, Jr., Louis

    1986-01-01

    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. Safety apparatus for nuclear reactor to prevent structural damage from overheating by core debris

    DOEpatents

    Gabor, John D.; Cassulo, John C.; Pedersen, Dean R.; Baker Jr., Louis

    1986-07-01

    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.

  4. Safety apparatus for nuclear reactor to prevent structural damage from overheating by core debris

    DOEpatents

    Gabor, J.D.; Cassulo, J.C.; Pedersen, D.R.; Baker, L. Jr.

    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 can be discharged from the reactor core. The invention provides a porous bed of 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.

  5. Status and problems of fusion reactor development.

    PubMed

    Schumacher, U

    2001-03-01

    Thermonuclear fusion of deuterium and tritium constitutes an enormous potential for a safe, environmentally compatible and sustainable energy supply. The fuel source is practically inexhaustible. Further, the safety prospects of a fusion reactor are quite favourable due to the inherently self-limiting fusion process, the limited radiologic toxicity and the passive cooling property. Among a small number of approaches, the concept of toroidal magnetic confinement of fusion plasmas has achieved most impressive scientific and technical progress towards energy release by thermonuclear burn of deuterium-tritium fuels. The status of thermonuclear fusion research activity world-wide is reviewed and present solutions to the complicated physical and technological problems are presented. These problems comprise plasma heating, confinement and exhaust of energy and particles, plasma stability, alpha particle heating, fusion reactor materials, reactor safety and environmental compatibility. The results and the high scientific level of this international research activity provide a sound basis for the realisation of the International Thermonuclear Experimental Reactor (ITER), whose goal is to demonstrate the scientific and technological feasibility of a fusion energy source for peaceful purposes.

  6. Impact of Active Control on Passive Safety Response Characteristics of Sodium-cooled Fast Reactors: I - Theoretical background

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Passerini, Stefano; Ponciroli, Roberto; Vilim, Richard B.

    Here, the interaction of the active control system with passive safety behavior is investigated for sodium-cooled fast reactors. A claim often made of advanced reactors is that they are passively safe against unprotected upset events. In practice, such upset events are not analyzed in the context of the plant control system, but rather the analyses are performed without considering the normally programmed response of the control system (open-loop approach). This represents an oversimplification of the safety case. The issue of passive safety override arises since the control system commands actuators whose motions have safety consequences. Depending on the upset involvingmore » the control system ( operator error, active control system failure, or inadvertent control system override), an actuator does not necessarily go in the same direction as needed for safety. So neglecting to account for control system action during an unprotected upset is nonconservative from a safety standpoint. It is important then, during the design of the plant, to consider the potential for the control system to work against the inherent and safe regulating effects of purposefully engineered temperature feedbacks.« less

  7. Impact of Active Control on Passive Safety Response Characteristics of Sodium-cooled Fast Reactors: I - Theoretical background

    DOE PAGES

    Passerini, Stefano; Ponciroli, Roberto; Vilim, Richard B.

    2017-06-21

    Here, the interaction of the active control system with passive safety behavior is investigated for sodium-cooled fast reactors. A claim often made of advanced reactors is that they are passively safe against unprotected upset events. In practice, such upset events are not analyzed in the context of the plant control system, but rather the analyses are performed without considering the normally programmed response of the control system (open-loop approach). This represents an oversimplification of the safety case. The issue of passive safety override arises since the control system commands actuators whose motions have safety consequences. Depending on the upset involvingmore » the control system ( operator error, active control system failure, or inadvertent control system override), an actuator does not necessarily go in the same direction as needed for safety. So neglecting to account for control system action during an unprotected upset is nonconservative from a safety standpoint. It is important then, during the design of the plant, to consider the potential for the control system to work against the inherent and safe regulating effects of purposefully engineered temperature feedbacks.« less

  8. Safety analysis

    NASA Technical Reports Server (NTRS)

    Knight, John C.

    1995-01-01

    We are engaged in a research program in safety-critical computing that is based on two case studies. We use these case studies to provide application-specific details of the various research issues, and as targets for evaluation of research ideas. The first case study is the Magnetic Stereotaxis System (MSS), an investigational device for performing human neurosurgery being developed in a joint effort between the Department of Physics at the University of Virginia and the Department of Neurosurgery at the University of Iowa. The system operates by manipulating a small permanent magnet (known as a 'seed') within the brain using an externally applied magnetic field. By varying the magnitude and gradient of the external magnetic field, the seed can be moved along a non-linear path and positioned at a site requiring therapy, e.g., a tumor. The magnetic field required for movement through brain tissue is extremely high, and is generated by a set of six superconducting magnets located in a housing surrounding the patient's head. The system uses two X-ray cameras positioned at right angles to detect in real time the locations of the seed and of X-ray opaque markers affixed to the patient's skull. the X-ray images are used to locate the objects of interest in a canonical frame of reference. the second case study is the University of Virginia Research Nuclear Reactor (UVAR). It is a 2 MW thermal, concrete-walled pool reactor. The system operates using 20 to 25 plate-type fuel assemblies placed on a rectangular grid plate. There are three scramable safety rods, and one non-scramable regulating rod that can be put in automatic mode. It was originally constructed in 1959 as a 1 MW system, and it was upgraded to 2 MW in 1973. Though only a research reactor rather than a power reactor, the issues raised are significant and can be related to the problems faced by full-scale reactor systems.

  9. Technology, safety, and costs of decommissioning reference nuclear research and test reactors: sensitivity of decommissioning radiation exposure and costs to selected parameters

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Konzek, G.J.

    1983-07-01

    Additional analyses of decommissioning at the reference research and test (R and T) reactors and analyses of five recent reactor decommissionings are made that examine some parameters not covered in the initial study report (NUREG/CR-1756). The parameters examined for decommissioning are: (1) the effect on costs and radiation exposure of plant size and/or type; (2) the effects on costs of increasing disposal charges and of unavailability of waste disposal capacity at licensed waste disposal facilities; and (3) the costs of and the available alternatives for the disposal of nuclear R and T reactor fuel assemblies.

  10. Operator Support System Design forthe Operation of RSG-GAS Research Reactor

    NASA Astrophysics Data System (ADS)

    Santoso, S.; Situmorang, J.; Bakhri, S.; Subekti, M.; Sunaryo, G. R.

    2018-02-01

    The components of RSG-GAS main control room are facing the problem of material ageing and technology obsolescence as well, and therefore the need for modernization and refurbishment are essential. The modernization in control room can be applied on the operator support system which bears the function in providing information for assisting the operator in conducting diagnosis and actions. The research purpose is to design an operator support system for RSG-GAS control room. The design was developed based on the operator requirement in conducting task operation scenarios and the reactor operation characteristics. These scenarios include power operation, low power operation and shutdown/scram reactor. The operator support system design is presented in a single computer display which contains structure and support system elements e.g. operation procedure, status of safety related components and operational requirements, operation limit condition of parameters, alarm information, and prognosis function. The prototype was developed using LabView software and consisted of components structure and features of the operator support system. Information of each component in the operator support system need to be completed before it can be applied and integrated in the RSG-GAS main control room.

  11. Preliminary safety analysis of Pb-Bi cooled 800 MWt modified CANDLE burn-up scheme based fast reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Su'ud, Zaki, E-mail: szaki@fi.itba.c.id; Sekimoto, H., E-mail: hsekimot@gmail.com

    2014-09-30

    Pb-Bi Cooled fast reactors with modified CANDLE burn-up scheme with 10 regions and 10 years cycle length has been investigated from neutronic aspects. In this study the safety aspect of such reactors have been investigated and discussed. Several condition of unprotected loss of flow (ULOF) and unprotected rod run-out transient over power (UTOP) have been simulated and the results show that the reactors excellent safety performance. At 80 seconds after unprotected loss of flow condition, the core flow rate drop to about 25% of its initial flow and slowly move toward its natural circulation level. The maximum fuel temperature canmore » be managed below 1000°C and the maximum cladding temperature can be managed below 700°C. The dominant reactivity feedback is radial core expansion and Doppler effect, followed by coolant density effect and fuel axial expansion effect.« less

  12. Current and anticipated uses of thermal hydraulic codes at the Japan Atomic Energy Research Institute

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Akimoto, Hajime; Kukita; Ohnuki, Akira

    1997-07-01

    The Japan Atomic Energy Research Institute (JAERI) is conducting several research programs related to thermal-hydraulic and neutronic behavior of light water reactors (LWRs). These include LWR safety research projects, which are conducted in accordance with the Nuclear Safety Commission`s research plan, and reactor engineering projects for the development of innovative reactor designs or core/fuel designs. Thermal-hydraulic and neutronic codes are used for various purposes including experimental analysis, nuclear power plant (NPP) safety analysis, and design assessment.

  13. Renewing Liquid Fueled Molten Salt Reactor Research and Development

    NASA Astrophysics Data System (ADS)

    Towell, Rusty; NEXT Lab Team

    2016-09-01

    Globally there is a desperate need for affordable, safe, and clean energy on demand. More than anything else, this would raise the living conditions of those in poverty around the world. An advanced reactor that utilizes liquid fuel and molten salts is capable of meeting these needs. Although, this technology was demonstrated in the Molten Salt Reactor Experiment (MSRE) at ORNL in the 60's, little progress has been made since the program was cancelled over 40 years ago. A new research effort has been initiated to advance the technical readiness level of key reactor components. This presentation will explain the motivation and initial steps for this new research initiative.

  14. Research reports (Annual reports). State: end of 1974

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    None

    1975-05-01

    This compilation of research reports is the third one to be published once a year in the frame of a comprehensive reporting on current investigations with regard to reactor safety. There are three types of reports: RS Research Reports, LRA Research Reports, GFK Research Reports. The RS Research Reports and the LRA Research Reports give information on the investigations sponsored by the Bundesminister fuer Forschung und Technologie (BMFT) and partly by the Bundesminister des Innern (BMI [SR 100, At T 85 a]) as individual reactor safety research projects. The GFK Research Reports inform about theoretical and experimental investigations on reactormore » safety conducted by the Gesellschaft fuer Kernforschung mbH (GFK), Karlsruhe. The Laboratorium fuer Reaktorregelung und Anlagensicherung (LRA), Muenchen-Garching, executes nine individual research projects comprehended under number At T 85 a. The work carried out by the GFK is included in the main project 'Nuclear Safety' (PNS). The single reports are attached to the main parts and focal points of the Research Program Reactor Safety. Therefore, at the head of the reports, under 'Project Number', not only the RS-, LRA- or GFK-Number but also the number of the main part of the Research Program which the reported investigation contributes to is noted. (orig.)« less

  15. Optimization of coupled multiphysics methodology for safety analysis of pebble bed modular reactor

    NASA Astrophysics Data System (ADS)

    Mkhabela, Peter Tshepo

    The research conducted within the framework of this PhD thesis is devoted to the high-fidelity multi-physics (based on neutronics/thermal-hydraulics coupling) analysis of Pebble Bed Modular Reactor (PBMR), which is a High Temperature Reactor (HTR). The Next Generation Nuclear Plant (NGNP) will be a HTR design. The core design and safety analysis methods are considerably less developed and mature for HTR analysis than those currently used for Light Water Reactors (LWRs). Compared to LWRs, the HTR transient analysis is more demanding since it requires proper treatment of both slower and much longer transients (of time scale in hours and days) and fast and short transients (of time scale in minutes and seconds). There is limited operation and experimental data available for HTRs for validation of coupled multi-physics methodologies. This PhD work developed and verified reliable high fidelity coupled multi-physics models subsequently implemented in robust, efficient, and accurate computational tools to analyse the neutronics and thermal-hydraulic behaviour for design optimization and safety evaluation of PBMR concept The study provided a contribution to a greater accuracy of neutronics calculations by including the feedback from thermal hydraulics driven temperature calculation and various multi-physics effects that can influence it. Consideration of the feedback due to the influence of leakage was taken into account by development and implementation of improved buckling feedback models. Modifications were made in the calculation procedure to ensure that the xenon depletion models were accurate for proper interpolation from cross section tables. To achieve this, the NEM/THERMIX coupled code system was developed to create the system that is efficient and stable over the duration of transient calculations that last over several tens of hours. Another achievement of the PhD thesis was development and demonstration of full-physics, three-dimensional safety analysis

  16. Herbal supplements: Research findings and safety.

    PubMed

    Pruitt, Rosanne; Lemanski, Ashley; Carroll, Adam

    2018-05-17

    Herbal supplements are used extensively worldwide without much awareness regarding their safety and efficacy. Extensive research to determine the safety, utility, and level of research support for commonly used herbs has culminated in an easily accessible summary chart for NP providers.

  17. Fast-acting nuclear reactor control device

    DOEpatents

    Kotlyar, Oleg M.; West, Phillip B.

    1993-01-01

    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.

  18. Dismantling the nuclear research reactor Thetis

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Michiels, P.

    The research reactor Thetis, in service since 1967 and stopped in 2003, is part of the laboratories of the institution of nuclear science of the University of Ghent. The reactor, of the pool-type, was used as a neutron-source for the production of radio-isotopes and for activation analyses. The reactor is situated in a water pool with inner diameter of 3 m. and a depth of 7.5 m. The reactor core is situated 5.3 m under water level. Besides the reactor, the pool contains pneumatic loops, handling tools, graphite blocks for neutron moderation and other experimental equipment. The building houses storagemore » rooms for fissile material and sources, a pneumatic circuit for transportation of samples, primary and secondary cooling circuits, water cleaning resin circuits, a ventilation system and other necessary devices. Because of the experimental character of the reactor, laboratories with glove boxes and other tools were needed and are included in the dismantling program. The building is in 3 levels with a crawl-space. The ground-floor contains the ventilation installation, the purification circuits with tanks, cooling circuits and pneumatic transport system. On the first floor, around the reactor hall, the control-room, visiting area, end-station for pneumatic transport, waste-storage room, fuel storage room and the labs are located. The second floor contains a few laboratories and end stations of the two high speed transfer tubes. The lowest level of the pool is situated under ground level. The reactor has been operated at a power of 150 kW and had a max operating power of 250 kW. Belgoprocess has been selected to decommission the reactor, the labs, storage halls and associated circuits to free release the building for conventional reuse and for the removal of all its internals as legal defined. Besides the dose-rate risk and contamination risk, there is also an asbestos risk of contamination. During construction of the installation, asbestos-containing materials

  19. Science and Safety: 'Acceptable' Risk

    ERIC Educational Resources Information Center

    Science News, 1976

    1976-01-01

    Stresses ways to answer questions related to widespread publicity - are nuclear reactors safe, will dangerous research in genetic manipulation be banned? - with emphasis on true meaning of safety as related to risks. (EB)

  20. Safety evaluation for packaging (onsite) plutonium recycle test reactor graphite cask

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Romano, T.

    This safety evaluation for packaging (SEP) provides the evaluation necessary to demonstrate that the Plutonium Recycle Test Reactor (PRTR) Graphite Cask meets the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B, fissile, non-highway route controlled quantities of radioactive material within the 300 Area of the Hanford Site. The scope of this SEP includes risk, shieldling, criticality, and.tiedown analyses to demonstrate that onsite transportation safety requirements are satisfied. This SEP also establishes operational and maintenance guidelines to ensure that transport of the PRTR Graphite Cask is performed safely in accordance with WHC-CM-2-14. This SEP is validmore » until October 1, 1999. After this date, an update or upgrade to this document is required.« less

  1. Needs and Requirements for Future Research Reactors (ORNL Perspectives)

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Ilas, Germina; Bryan, Chris; Gehin, Jess C.

    2016-02-10

    The High Flux Isotope Reactor (HFIR) is a vital national and international resource for neutron science research, production of radioisotopes, and materials irradiation. While HFIR is expected to continue operation for the foreseeable future, interest is growing in understanding future research reactors features, needs, and requirements. To clarify, discuss, and compile these needs from the perspective of Oak Ridge National Laboratory (ORNL) research and development (R&D) missions, a workshop, titled “Needs and Requirements for Future Research Reactors”, was held at ORNL on May 12, 2015. The workshop engaged ORNL staff that is directly involved in research using HFIR to collectmore » valuable input on the reactor’s current and future missions. The workshop provided an interactive forum for a fruitful exchange of opinions, and included a mix of short presentations and open discussions. ORNL staff members made 15 technical presentations based on their experience and areas of expertise, and discussed those capabilities of the HFIR and future research reactors that are essential for their current and future R&D needs. The workshop was attended by approximately 60 participants from three ORNL directorates. The agenda is included in Appendix A. This document summarizes the feedback provided by workshop contributors and participants. It also includes information and insights addressing key points that originated from the dialogue started at the workshop. A general overview is provided on the design features and capabilities of high performance research reactors currently in use or under construction worldwide. Recent and ongoing design efforts in the US and internationally are briefly summarized, followed by conclusions and recommendations.« less

  2. Analysis of Accidents at the Pakistan Research Reactor-1 Using Proposed Mixed-Fuel (HEU and LEU) Core

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Bokhari, Ishtiaq H.

    2004-12-15

    The Pakistan Research Reactor-1 (PARR-1) was converted from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel in 1991. The reactor is running successfully, with an upgraded power level of 10 MW. To save money on the purchase of costly fresh LEU fuel elements, the use of less burnt HEU spent fuel elements along with the present LEU fuel elements is being considered. The proposal calls for the HEU fuel elements to be placed near the thermal column to gain the required excess reactivity. In the present study the safety analysis of a proposed mixed-fuel core has been carried outmore » at a calculated steady-state power level of 9.8 MW. Standard computer codes and correlations were employed to compute various parameters. Initiating events in reactivity-induced accidents involve various modes of reactivity insertion, namely, start-up accident, accidental drop of a fuel element on the core, flooding of a beam tube with water, and removal of an in-pile experiment during reactor operation. For each of these transients, time histories of reactor power, energy released, temperature, and reactivity were determined.« less

  3. Aircraft fire safety research

    NASA Technical Reports Server (NTRS)

    Botteri, Benito P.

    1987-01-01

    During the past 15 years, very significant progress has been made toward enhancing aircraft fire safety in both normal and hostile (combat) operational environments. Most of the major aspects of the aircraft fire safety problem are touched upon here. The technology of aircraft fire protection, although not directly applicable in all cases to spacecraft fire scenarios, nevertheless does provide a solid foundation to build upon. This is particularly true of the extensive research and testing pertaining to aircraft interior fire safety and to onboard inert gas generation systems, both of which are still active areas of investigation.

  4. Chemical Safety for Sustainability: Research Action Plan

    EPA Pesticide Factsheets

    The Strategic Research Action Plan for EPA’s Chemical Safety for Sustainability research program presents the purpose, design and themes of the Agency’s research efforts to ensure safety in the design, manufacture and use of existing and future chemicals.

  5. Further Development of Crack Growth Detection Techniques for US Test and Research Reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Kohse, Gordon; Carpenter, David M.; Ostrovsky, Yakov

    One of the key issues facing Light Water Reactors (LWRs) in extending lifetimes beyond 60 years is characterizing the combined effect of irradiation and water chemistry on material degradation and failure. Irradiation Assisted Stress Corrosion Cracking (IASCC), in which a crack propagates in a susceptible material under stress in an aggressive environment, is a mechanism of particular concern. Full understanding of IASCC depends on real time crack growth data acquired under relevant irradiation conditions. Techniques to measure crack growth in actively loaded samples under irradiation have been developed outside the US - at the Halden Boiling Water Reactor, for example.more » Several types of IASCC tests have also been deployed at the MITR, including passively loaded crack growth measurements and actively loaded slow strain rate tests. However, there is not currently a facility available in the US to measure crack growth on actively loaded, pre-cracked specimens in LWR irradiation environments. A joint program between the Idaho National Laboratory (INL) and the Massachusetts Institute of Technology (MIT) Nuclear Reactor Laboratory (NRL) is currently underway to develop and demonstrate such a capability for US test and research reactors. Based on the Halden design, the samples will be loaded using miniature high pressure bellows and a compact loading mechanism, with crack length measured in real time using the switched Direct Current Potential Drop (DCPD) method. The basic design and initial mechanical testing of the load system and implementation of the DCPD method have been previously reported. This paper presents the results of initial autoclave testing at INL and the adaptation of the design for use in the high pressure, high temperature water loop at the MITR 6 MW research reactor, where an initial demonstration is planned in mid-2015. Materials considerations for the high pressure bellows are addressed. Design modifications to the loading mechanism required

  6. Neutronic and thermal-hydraulic analysis of fission molybdenum-99 production at Tehran Research Reactor using LEU plate targets.

    PubMed

    Abedi, Ebrahim; Ebrahimkhani, Marzieh; Davari, Amin; Mirvakili, Seyed Mohammad; Tabasi, Mohsen; Maragheh, Mohammad Ghannadi

    2016-12-01

    Efficient and safe production of molybdenum-99 ( 99 Mo) radiopharmaceutical at Tehran Research Reactor (TRR) via fission of LEU targets is studied. Neutronic calculations are performed to evaluate produced 99 Mo activity, core neutronic safety parameters and also the power deposition values in target plates during a 7 days irradiation interval. Thermal-hydraulic analysis has been also carried out to obtain thermal behavior of these plates. Using Thermal-hydraulic analysis, it can be concluded that the safety parameters are satisfied in the current study. Consequently, the present neutronic and thermal-hydraulic calculations show efficient 99 Mo production is accessible at significant activity values in TRR current core configuration. Copyright © 2016 Elsevier Ltd. All rights reserved.

  7. JPL in-house fluidized-bed reactor research

    NASA Technical Reports Server (NTRS)

    Rohatgi, N. K.

    1984-01-01

    Fluidized bed reactor research techniques for fabrication of quartz linears was reviewed. Silane pyrolysis was employed in this fabrication study. Metallic contaminant levels in the silicon particles were below levels detectable by emission spectroscopy.

  8. Light Water Reactor Sustainability Program: Risk-Informed Safety Margins Characterization (RISMC) Pathway Technical Program Plan

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Smith, Curtis; Rabiti, Cristian; Martineau, Richard

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

  9. Proceedings of a Symposium on Advanced Compact Reactor Systems

    NASA Technical Reports Server (NTRS)

    1983-01-01

    Reactor system technologies suitable for a variety of aerospace and terrestrial applications are considered. Technologies, safety and regulatory considerations, potential applications, and research and development opportunities are covered.

  10. Integrated Decision-Making Tool to Develop Spent Fuel Strategies for Research Reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Beatty, Randy L; Harrison, Thomas J

    IAEA Member States operating or having previously operated a Research Reactor are responsible for the safe and sustainable management and disposal of associated radioactive waste, including research reactor spent nuclear fuel (RRSNF). This includes the safe disposal of RRSNF or the corresponding equivalent waste returned after spent fuel reprocessing. One key challenge to developing general recommendations lies in the diversity of spent fuel types, locations and national/regional circumstances rather than mass or volume alone. This is especially true given that RRSNF inventories are relatively small, and research reactors are rarely operated at a high power level or duration typical ofmore » commercial power plants. Presently, many countries lack an effective long-term policy for managing RRSNF. This paper presents results of the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) #T33001 on Options and Technologies for Managing the Back End of the Research Reactor Nuclear Fuel Cycle which includes an Integrated Decision Making Tool called BRIDE (Back-end Research reactor Integrated Decision Evaluation). This is a multi-attribute decision-making tool that combines the Total Estimated Cost of each life-cycle scenario with Non-economic factors such as public acceptance, technical maturity etc and ranks optional back-end scenarios specific to member states situations in order to develop a specific member state strategic plan with a preferred or recommended option for managing spent fuel from Research Reactors.« less

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Snoj, L.; Sklenka, L.; Rataj, J.

    2012-07-01

    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 differentmore » 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)« less

  12. Current and prospective safety issues at the HFBR

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Tichler, P.R.

    The Brookhaven high-flux beam reactor (HFBR) was designed primarily to produce external neutron beams for experimental research. It is cooled, moderated, and reflected by heavy water and uses materials test reactor and engineering test reactor type of fuel elements containing enriched uranium. The reactor power when operation began in 1965 was 40 MW, was raised to 60 MW in 1982 after a number of plant modifications, and operated at that level until 1989. Since that time, safety questions have been raised that resulted in extended shutdowns and a reduction in operating power to 30 MW. This paper discusses the principalmore » safety issues and plans for their resolution and return to 60-MW operation. In addition, radiation embrittlement of the reactor vessel and thermal shield and its effect on the life of the facility are briefly discussed.« less

  13. Manned space flight nuclear system safety. Volume 3: Reactor system preliminary nuclear safety analysis. Part 3: Nuclear Safety Analysis Document (NSAD)

    NASA Technical Reports Server (NTRS)

    1972-01-01

    Nuclear safety analysis as applied to a space base mission is presented. The nuclear safety analysis document summarizes the mission and the credible accidents/events which may lead to nuclear hazards to the general public. The radiological effects and associated consequences of the hazards are discussed in detail. The probability of occurrence is combined with the potential number of individuals exposed to or above guideline values to provide a measure of accident and total mission risk. The overall mission risk has been determined to be low with the potential exposure to or above 25 rem limited to less than 4 individuals per every 1000 missions performed. No radiological risk to the general public occurs during the prelaunch phase at KSC. The most significant risks occur from prolonged exposure to reactor debris following land impact generally associated with the disposal phase of the mission where fission product inventories can be high.

  14. Conflicts of interest in vaccine safety research.

    PubMed

    DeLong, Gayle

    2012-01-01

    Conflicts of interest (COIs) cloud vaccine safety research. Sponsors of research have competing interests that may impede the objective study of vaccine side effects. Vaccine manufacturers, health officials, and medical journals may have financial and bureaucratic reasons for not wanting to acknowledge the risks of vaccines. Conversely, some advocacy groups may have legislative and financial reasons to sponsor research that finds risks in vaccines. Using the vaccine-autism debate as an illustration, this article details the conflicts of interest each of these groups faces, outlines the current state of vaccine safety research, and suggests remedies to address COIs. Minimizing COIs in vaccine safety research could reduce research bias and restore greater trust in the vaccine program.

  15. RERTR 2009 (Reduced Enrichment for Research and Test Reactors)

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Totev, T.; Stevens, J.; Kim, Y. S.

    2010-03-01

    The U.S. Department of Energy/National Nuclear Security Administration's Office of Global Threat Reduction in cooperation with the China Atomic Energy Authority and International Atomic Energy Agency hosted the 'RERTR 2009 International Meeting on Reduced Enrichment for Research and Test Reactors.' The meeting was organized by Argonne National Laboratory, China Institute of Atomic Energy and Idaho National Laboratory and was held in Beijing, China from November 1-5, 2009. This was the 31st annual meeting in a series on the same general subject regarding the conversion of reactors within the Global Threat Reduction Initiative (GTRI). The Reduced Enrichment for Research and Testmore » Reactors (RERTR) Program develops technology necessary to enable the conversion of civilian facilities using high enriched uranium (HEU) to low enriched uranium (LEU) fuels and targets.« less

  16. Yale High Energy Physics Research: Precision Studies of Reactor Antineutrinos

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Heeger, Karsten M.

    2014-09-13

    This report presents experimental research at the intensity frontier of particle physics with particular focus on the study of reactor antineutrinos and the precision measurement of neutrino oscillations. The experimental neutrino physics group of Professor Heeger and Senior Scientist Band at Yale University has had leading responsibilities in the construction and operation of the Daya Bay Reactor Antineutrino Experiment and made critical contributions to the discovery of non-zeromore » $$\\theta_{13}$$. Heeger and Band led the Daya Bay detector management team and are now overseeing the operations of the antineutrino detectors. Postdoctoral researchers and students in this group have made leading contributions to the Daya Bay analysis including the prediction of the reactor antineutrino flux and spectrum, the analysis of the oscillation signal, and the precision determination of the target mass yielding unprecedented precision in the relative detector uncertainty. Heeger's group is now leading an R\\&D effort towards a short-baseline oscillation experiment, called PROSPECT, at a US research reactor and the development of antineutrino detectors with advanced background discrimination.« less

  17. NUCLEAR REACTOR

    DOEpatents

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  18. Thermal-hydraulic interfacing code modules for CANDU reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Liu, W.S.; Gold, M.; Sills, H.

    1997-07-01

    The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.

  19. Using the Human Systems Simulation Laboratory at Idaho National Laboratory for Safety Focused Research

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Joe, Jeffrey .C; Boring, Ronald L.

    Under the United States (U.S.) Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) program, researchers at Idaho National Laboratory (INL) have been using the Human Systems Simulation Laboratory (HSSL) to conduct critical safety focused Human Factors research and development (R&D) for the nuclear industry. The LWRS program has the overall objective to develop the scientific basis to extend existing nuclear power plant (NPP) operating life beyond the current 60-year licensing period and to ensure their long-term reliability, productivity, safety, and security. One focus area for LWRS is the NPP main control room (MCR), because many of the instrumentation andmore » control (I&C) system technologies installed in the MCR, while highly reliable and safe, are now difficult to replace and are therefore limiting the operating life of the NPP. This paper describes how INL researchers use the HSSL to conduct Human Factors R&D on modernizing or upgrading these I&C systems in a step-wise manner, and how the HSSL has addressed a significant gap in how to upgrade systems and technologies that are built to last, and therefore require careful integration of analog and new advanced digital technologies.« less

  20. Keeping nurse researchers safe: workplace health and safety issues.

    PubMed

    Barr, Jennieffer; Welch, Anthony

    2012-07-01

    This article is a report of a qualitative study of workplace health and safety issues in nursing research. Researcher health and safety have become increasing concerns as there is an increased amount of research undertaken in the community and yet there is a lack of appropriate guidelines on how to keep researchers safe when undertaking fieldwork. This study employed a descriptive qualitative approach, using different sources of data to find any references to researcher health and safety issues. A simple descriptive approach to inquiry was used for this study. Three approaches to data collection were used: interviews with 15 researchers, audits of 18 ethics applications, and exploration of the literature between 1992 and 2010 for examples of researcher safety issues. Data analysis from the three approaches identified participant comments, narrative descriptions or statements focused on researcher health and safety. Nurse researchers' health and safety may be at risk when conducting research in the community. Particular concern involves conducting sensitive research where researchers are physically at risk of being harmed, or being exposed to the development of somatic symptoms. Nurse researchers may perceive the level of risk of harm as lower than the actual or potential harm present in research. Nurse researchers do not consistently implement risk assessment before and during research. Researcher health and safety should be carefully considered at all stages of the research process. Research focusing on sensitive data and vulnerable populations need to consider risk minimization through strategies such as appropriate researcher preparation, safety during data collection, and debriefing if required. © 2012 Blackwell Publishing Ltd.

  1. Steady state and LOCA analysis of Kartini reactor using RELAP5/SCDAP code: The role of passive system

    NASA Astrophysics Data System (ADS)

    Antariksawan, Anhar R.; Wahyono, Puradwi I.; Taxwim

    2018-02-01

    Safety is the priority for nuclear installations, including research reactors. On the other hand, many studies have been done to validate the applicability of nuclear power plant based best estimate computer codes to the research reactor. This study aims to assess the applicability of the RELAP5/SCDAP code to Kartini research reactor. The model development, steady state and transient due to LOCA calculations have been conducted by using RELAP5/SCDAP. The calculation results are compared with available measurements data from Kartini research reactor. The results show that the RELAP5/SCDAP model steady state calculation agrees quite well with the available measurement data. While, in the case of LOCA transient simulations, the model could result in reasonable physical phenomena during the transient showing the characteristics and performances of the reactor against the LOCA transient. The role of siphon breaker hole and natural circulation in the reactor tank as passive system was important to keep reactor in safe condition. It concludes that the RELAP/SCDAP could be use as one of the tool to analyse the thermal-hydraulic safety of Kartini reactor. However, further assessment to improve the model is still needed.

  2. The present situations and perspectives on utilization of research reactors in Thailand

    NASA Astrophysics Data System (ADS)

    Chongkum, Somporn

    2002-01-01

    The Thai Research Reactor 1/Modification 1, a TRIGA Mark III reactor, went critical on November 7, 1977. It has been playing a central role in the development of both Office of Atomic Energy for Peace (OAEP) and nuclear application in Thailand. It has a maximum power of 2 MW (thermal) at steady state and a pulsing capacity of 2000 MW. The highest thermal neutron flux at a central thimber is 1×10 13 n/cm 2/s, which is extensively utilized for radioisotope production, neutron activation analysis and neutron beam experiments, i.e. neutron scattering, prompt gamma analysis and neutron radiography. Following the nuclear technological development, the OAEP is in the process of establishing the Ongkharak Nuclear Research Center (ONRC). The center is being built in Nakhon Nayok province, 60 km northeast of Bangkok. The centerpiece of the ONRC is a multipurpose 10 MW TRIGA research reactor. Facilities are included for the production of radioisotopes for medicine, industry and agriculture, neutron transmutation doping of silicon, and neutron capture therapy. The neutron beam facilities will also be utilized for applied research and technology development as well as training in reactor operations, performance of experiments and reactor physics. This paper describes a recent program of utilization as well as a new research reactor for enlarging the perspectives of its utilization in the future.

  3. Training courses on neutron detection systems on the ISIS research reactor: on-site and through internet training

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Lescop, B.; Badeau, G.; Ivanovic, S.

    Today, ISIS research reactor is an essential tool for Education and Training programs organized by the National Institute for Nuclear Science and Technology (INSTN) from CEA. In the field of nuclear instrumentation, the INSTN offers both, theoretical courses and training courses on the use of neutron detection systems taking advantage of the ISIS research reactor for the supply of a wide range of neutron fluxes. This paper describes the content of the training carried out on the use of neutron detectors and detection systems, on-site or remote. The ISIS reactor is a 700 kW open core pool type reactor. Themore » facility is very flexible since neutron detectors can be inserted into the core or its vicinity, and be used at different levels of power according to the needs of the course. Neutron fluxes, typically ranging from 1 to 10{sup 12} n/cm{sup 2}.s, can be obtained for the characterisation of the neutron detectors and detection systems. For the monitoring of the neutron density at low level of power, the Instrumentation and Control (I and C) system of the reactor is equipped with two detection systems, named BN1 and BN2. Each way contains a fission chamber, type CFUL01, connected to an electronic system type SIREX.The system works in pulse mode and exhibits two outputs: the counting rate and the doubling time. For the high level of power, the I and C is equipped with two detection systems HN1 and HN2.Each way contain a boron ionization chamber (type CC52) connected to an electronics system type SIREX. The system works in current mode and has two outputs: the current and the doubling time. For each mode, the trainees can observe and measure the signal at the different stages of the electronic system, with an oscilloscope. They can understand the role of each component of the detection system: detector, cable and each electronic block. The limitation of the detection modes and their operating range can be established from the measured signal. The trainees can

  4. Criticality Safety Evaluation of the LLNL Inherently Safe Subcritical Assembly (ISSA)

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Percher, Catherine

    2012-06-19

    The LLNL Nuclear Criticality Safety Division has developed a training center to illustrate criticality safety and reactor physics concepts through hands-on experimental training. The experimental assembly, the Inherently Safe Subcritical Assembly (ISSA), uses surplus highly enriched research reactor fuel configured in a water tank. The training activities will be conducted by LLNL following the requirements of an Integration Work Sheet (IWS) and associated Safety Plan. Students will be allowed to handle the fissile material under the supervision of LLNL instructors. This report provides the technical criticality safety basis for instructional operations with the ISSA experimental assembly.

  5. 75 FR 12554 - Mine Safety and Health Research Advisory Committee, National Institute for Occupational Safety...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-03-16

    ... DEPARTMENT OF HEALTH AND HUMAN SERVICES Centers for Disease Control and Prevention Mine Safety and Health Research Advisory Committee, National Institute for Occupational Safety and Health (MSHRAC, NIOSH... priorities in mine safety and health research, including grants and contracts for such research, 30 U.S.C...

  6. Impact induced response spectrum for the safety evaluation of the high flux isotope reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Chang, S.J.

    1997-05-01

    The dynamic impact to the nearby HFIR reactor vessel caused by heavy load drop is analyzed. The impact calculation is carried out by applying the ABAQUS computer code. An impact-induced response spectrum is constructed in order to evaluate whether the HFIR vessel and the shutdown mechanism may be disabled. For the frequency range less than 10 Hz, the maximum spectral velocity of impact is approximately equal to that of the HFIR seismic design-basis spectrum. For the frequency range greater than 10 Hz, the impact-induced response spectrum is shown to cause no effect to the control rod and the shutdown mechanism.more » An earlier seismic safety assessment for the HFIR control and shutdown mechanism was made by EQE. Based on EQE modal solution that is combined with the impact-induced spectrum, it is concluded that the impact will not cause any damage to the shutdown mechanism, even while the reactor is in operation. The present method suggests a general approach for evaluating the impact induced damage to the reactor by applying the existing finite element modal solution that has been carried out for the seismic evaluation of the reactor.« less

  7. Research reactor loading pattern optimization using estimation of distribution algorithms

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Jiang, S.; Ziver, K.; AMCG Group, RM Consultants, Abingdon

    2006-07-01

    A new evolutionary search based approach for solving the nuclear reactor loading pattern optimization problems is presented based on the Estimation of Distribution Algorithms. The optimization technique developed is then applied to the maximization of the effective multiplication factor (K{sub eff}) of the Imperial College CONSORT research reactor (the last remaining civilian research reactor in the United Kingdom). A new elitism-guided searching strategy has been developed and applied to improve the local convergence together with some problem-dependent information based on the 'stand-alone K{sub eff} with fuel coupling calculations. A comparison study between the EDAs and a Genetic Algorithm with Heuristicmore » Tie Breaking Crossover operator has shown that the new algorithm is efficient and robust. (authors)« less

  8. 75 FR 70042 - In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-11-16

    ... NUCLEAR REGULATORY COMMISSION [Docket Nos (Redacted), License Nos (Redacted), EA (Redacted); NRC- 2010-0351] In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent Nuclear Fuel; Order Modifying License (Effective Immediately) I. The licensees identified in...

  9. 75 FR 79423 - In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-12-20

    ... NUCLEAR REGULATORY COMMISSION [Docket Nos. (Redacted), License Nos.: (Redacted), EA (Redacted); NRC- 2010-0351] In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent Nuclear Fuel; Order Modifying License (Effective Immediately) I The licensees identified in...

  10. 78 FR 26811 - Dow Chemical Company, Dow TRIGA Research Reactor; License Renewal for the Dow Chemical TRIGA...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-05-08

    ... Research Reactor; License Renewal for the Dow Chemical TRIGA Research Reactor; Supplemental Information and... 20, 2012 (77 FR 42771), ``License Renewal for the Dow Chemical TRIGA Research Reactor,'' to inform... Chemical Company which would authorize continued operation of the Dow TRIGA Research Reactor. The notice...

  11. IEA-R1 Nuclear Research Reactor: 58 Years of Operating Experience and Utilization for Research, Teaching and Radioisotopes Production

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Cardenas, Jose Patricio Nahuel; Filho, Tufic Madi; Saxena, Rajendra

    IEA-R1 research reactor at the Instituto de Pesquisas Energeticas e Nucleares (Nuclear and Energy Research Institute) IPEN, Sao Paulo, Brazil is the largest power research reactor in Brazil, with a maximum power rating of 5 MWth. It is being used for basic and applied research in the nuclear and neutron related sciences, for the production of radioisotopes for medical and industrial applications, and for providing services of neutron activation analysis, real time neutron radiography, and neutron transmutation doping of silicon. IEA-R1 is a swimming pool reactor, with light water as the coolant and moderator, and graphite and beryllium as reflectors.more » The reactor was commissioned on September 16, 1957 and achieved its first criticality. It is currently operating at 4.5 MWth with a 60-hour cycle per week. In the early sixties, IPEN produced {sup 131}I, {sup 32}P, {sup 198}Au, {sup 24}Na, {sup 35}S, {sup 51}Cr and labeled compounds for medical use. During the past several years, a concerted effort has been made in order to upgrade the reactor power to 5 MWth through refurbishment and modernization programs. One of the reasons for this decision was to produce {sup 99}Mo at IPEN. The reactor cycle will be gradually increased to 120 hours per week continuous operation. It is anticipated that these programs will assure the safe and sustainable operation of the IEA-R1 reactor for several more years, to produce important primary radioisotopes {sup 99}Mo, {sup 125}I, {sup 131}I, {sup 153}Sm and {sup 192}Ir. Currently, all aspects of dealing with fuel element fabrication, fuel transportation, isotope processing, and spent fuel storage are handled by IPEN at the site. The reactor modernization program is slated for completion by 2015. This paper describes 58 years of operating experience and utilization of the IEA-R1 research reactor for research, teaching and radioisotopes production. (authors)« less

  12. Reversal of OFI and CHF in Research Reactors Operating at 1 to 50 Bar. Version 1.0

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Kalimullah, M.; Olson, A. P.; Dionne, B.

    2014-02-28

    The conditions at which the critical heat flux (CHF) and the heat flux at the onset of Ledinegg flow instability (OFI) are equal, are determined for a coolant channel with uniform heat flux as a function of five independent parameters: the channel exit pressure (P), heated length (Lh) , heated diameter (Dh), inlet temperature (Tin), and mass flux (G). A diagram is made by plotting the mass flux and heat flux at the OFI-CHF intersection (reversal from CHF > OFI to CHF < OFI as G increases) as a function of P (1 to 50 bar), for 36 combinations ofmore » the remaining three parameters (Lh , Dh , Tin): Lh = 0.28, 0.61, 1.18 m; Dh = 3, 4, 6, 8 mm; Tin = 30, 50, 70 °C. The use of the diagram to scope whether a research reactor is OFI-limited (below the curve) or CHF-limited based on the five parameters of its coolant channel is described. Justification for application of the diagram to research reactors with axially non-uniform heat flux is provided. Due to its limitations (uncertainties not included), the diagram cannot replace the detailed thermal-hydraulic analysis required for a reactor safety analysis. In order to make the OFI-CHF intersection diagram, two world-class CHF prediction methods (the Hall-Mudawar correlation and the extended Groeneveld 2006 table) are compared for 216 combinations of the five independent parameters. The two widely used OFI correlations (the Saha- Zuber and the Whittle-Forgan with η = 32.5) are also compared for the same combinations of the five parameters. The extended Groeneveld table and the Whittle-Forgan OFI correlation are selected for use in making the diagram. Using the above five design parameters, a research reactor can be represented by a point on the reversal diagram, and the diagram can be used to scope, without a thermal-hydraulic calculation, whether the OFI will occur before the CHF, or the CHF will occur before the OFI when the reactor power is increased keeping the five parameters fixed.« less

  13. Anthropology in Agricultural Health and Safety Research and Intervention.

    PubMed

    Arcury, Thomas

    2017-01-01

    Agriculture remains a dangerous industry, even as agricultural science and technology continue to advance. Research that goes beyond technological changes to address safety culture and policy are needed to improve health and safety in agriculture. In this commentary, I consider the potential for anthropology to contribute to agricultural health and safety research by addressing three aims: (1) I briefly consider what the articles in this issue of the Journal of Agromedicine say about anthropologists in agricultural health and safety; (2) I discuss what anthropologists can add to agricultural health and safety research; and (3) I examine ways in which anthropologists can participate in agricultural health and safety research. In using their traditions of rigorous field research to understand how those working in agriculture perceive and interpret factors affecting occupational health and safety (their "emic" perspective), and translating this perspective to improve the understanding of occupational health professionals and policy makers (an "etic" perspective), anthropologists can expose myths that limit improvements in agricultural health and safety. Addressing significant questions, working with the most vulnerable agricultural communities, and being outside establishment agriculture provide anthropologists with the opportunity to improve health and safety policy and regulation in agriculture.

  14. Safety management of complex research operators

    NASA Technical Reports Server (NTRS)

    Brown, W. J.

    1981-01-01

    Complex research and technology operations present varied potential hazards which are addressed in a disciplined, independent safety review and approval process. Potential hazards vary from high energy fuels to hydrocarbon fuels, high pressure systems to high voltage systems, toxic chemicals to radioactive materials and high speed rotating machinery to high powered lasers. A Safety Permit System presently covers about 600 potentially hazardous operations. The Safety Management Program described is believed to be a major factor in maintaining an excellent safety record.

  15. Safety survey report EBR-II safety survey, ANL-west health protection, industrial safety and fire protection survey

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Dunbar, K.A.

    1972-01-10

    A safety survey covering the disciplines of Reactor Safety, Nuclear Criticality Safety, Health Protection and Industrial Safety and Fire Protection was conducted at the ANL-West EBR-II FEF Complex during the period January 10-18, 1972. In addition, the entire ANL-West site was surveyed for Health Protection and Industrial Safety and Fire Protection. The survey was conducted by members of the AEC Chicago Operations Office, a member of RDT-HQ and a member of the RDT-ID site office. Eighteen recommendations resulted from the survey, eleven in the area of Industrial Safety and Fire Protection, five in the area of Reactor Safety and twomore » in the area of Nuclear Criticality Safety.« less

  16. Neutron beams implemented at nuclear research reactors for BNCT

    NASA Astrophysics Data System (ADS)

    Bavarnegin, E.; Kasesaz, Y.; Wagner, F. M.

    2017-05-01

    This paper presents a survey of neutron beams which were or are in use at 56 Nuclear Research Reactors (NRRs) in order to be used for BNCT, either for treatment or research purposes in aspects of various combinations of materials that were used in their Beam Shaping Assembly (BSA) design, use of fission converters and optimized beam parameters. All our knowledge about BNCT is indebted to researches that have been done in NRRs. The results of about 60 years research in BNCT and also the successes of this method in medical treatment of tumors show that, for the development of BNCT as a routine cancer therapy method, hospital-based neutron sources are needed. Achieving a physical data collection on BNCT neutron beams based on NRRs will be helpful for beam designers in developing a non-reactor based neutron beam.

  17. The MSFR as a flexible CR reactor: the viewpoint of safety

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Fiorina, C.; Cammi, A.; Franceschini, F.

    2013-07-01

    In this paper, the possibility has first been discussed of using the liquid-fuelled Molten Salt Fast Reactor (MSFR) as a flexible conversion ratio (CR) reactor without design modification. By tuning the reprocessing rate it is possible to determine the content of fission products in the core, which in turn can significantly affect the neutron economy without incurring in solubility problems. The MSFR can thus be operated as U-233 breeder (CR>1), iso-breeder (CR=1) and burner reactor (CR<1). In particular a 40 year doubling time can be achieved, as well as a considerable Transuranics and MA (minor actinide) burning rate equal tomore » about 150 kg{sub HN}/GWE-yr. The safety parameters of the MSFR have then been evaluated for different fuel cycle strategies. Th use and a softer spectrum combine to give a strong Doppler coefficient, one order of magnitude higher compared to traditional fast reactors (FRs). The fuel expansion coefficient is comparable to the Doppler coefficient and is only mildly affected by core compositions, thus assisting the fuel cycle flexibility of the MSFR. βeff and generation time are comparable to the case of traditional FRs, if a static fuel is assumed. A notable reduction of βeff is caused by salt circulation, but a low value of this parameter is a limited concern in the MSFR thanks to the lack of a burnup reactivity swing and of positive feedbacks. A simple approach has also been developed to evaluate the MSFR capabilities to withstand all typical double-fault accidents, for different fuel cycle options.« less

  18. Overview of safety research

    NASA Technical Reports Server (NTRS)

    Enders, J. H.

    1978-01-01

    Aircraft safety is reviewed by first establishing a perspective of air transportation accidents as a function of calendar year, geographic area, and phase of flight, and then by describing the threats to safety and NASA research underway in the three representative areas of engine operational problems, meteorological phenomena, and fire. Engine rotor burst protection, aircraft nacelle fire extinguishment, the aircraft-weather interface, severe weather wind shears and turbulence, clear air turbulence, and lightning are among the topics covered. Fire impact management through fire resistant materials technology development is emphasized.

  19. Nuclear safety considerations in the conceptual design of a fast reactor for space electric power and propulsion

    NASA Technical Reports Server (NTRS)

    Hsieh, T.-M.; Koenig, D. R.

    1977-01-01

    Some nuclear safety aspects of a 3.2 mWt heat pipe cooled fast reactor with out-of-core thermionic converters are discussed. Safety related characteristics of the design including a thin layer of B4C surrounding the core, the use of heat pipes and BeO reflector assembly, the elimination of fuel element bowing, etc., are highlighted. Potential supercriticality hazards and countermeasures are considered. Impacts of some safety guidelines of space transportation system are also briefly discussed, since the currently developing space shuttle would be used as the primary launch vehicle for the nuclear electric propulsion spacecraft.

  20. NASA's aviation safety research and technology program

    NASA Technical Reports Server (NTRS)

    Fichtl, G. H.

    1977-01-01

    Aviation safety is challenged by the practical necessity of compromising inherent factors of design, environment, and operation. If accidents are to be avoided these factors must be controlled to a degree not often required by other transport modes. The operational problems which challenge safety seem to occur most often in the interfaces within and between the design, the environment, and operations where mismatches occur due to ignorance or lack of sufficient understanding of these interactions. Under this report the following topics are summarized: (1) The nature of operating problems, (2) NASA aviation safety research, (3) clear air turbulence characterization and prediction, (4) CAT detection, (5) Measurement of Atmospheric Turbulence (MAT) Program, (6) Lightning, (7) Thunderstorm gust fronts, (8) Aircraft ground operating problems, (9) Aircraft fire technology, (10) Crashworthiness research, (11) Aircraft wake vortex hazard research, and (12) Aviation safety reporting system.

  1. Qualification of Simulation Software for Safety Assessment of Sodium Cooled Fast Reactors. Requirements and Recommendations

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Brown, Nicholas R.; Pointer, William David; Sieger, Matt

    2016-04-01

    The goal of this review is to enable application of codes or software packages for safety assessment of advanced sodium-cooled fast reactor (SFR) designs. To address near-term programmatic needs, the authors have focused on two objectives. First, the authors have focused on identification of requirements for software QA that must be satisfied to enable the application of software to future safety analyses. Second, the authors have collected best practices applied by other code development teams to minimize cost and time of initial code qualification activities and to recommend a path to the stated goal.

  2. 60th Anniversary of electricity production from light water reactors: Historical review of the contribution of materials science to the safety of the pressure vessel

    NASA Astrophysics Data System (ADS)

    van Duysen, J. C.; Meric de Bellefon, G.

    2017-02-01

    The first light water nuclear reactor dedicated to electricity production was commissioned in Shippingport, Pennsylvania in the United States in 1957. Sixty years after the event, it is clear that this type of reactor will be a major source of electricity and one of the key solutions to limit climate change in the 21st century. This article pays homage to the teams that contributed to this achievement by their involvement in research and development and their determination to push back the frontiers of knowledge. Via a few examples of scientific or technological milestones, it describes the evolution of ideas, models, and techniques during the last 60 years, and gives the current state-of-the-art in areas related to the safety of the reactor pressure vessel. Among other topics, it focuses on vessel manufacturing, steel fracture mechanics analysis, and understanding of irradiation-induced damage.

  3. Overview of Federal Motor Carrier Safety Administration safety training research for new entrant motor carriers : [research brief].

    DOT National Transportation Integrated Search

    2015-07-01

    In 2002, the Federal Motor Carrier Safety Administration (FMCSA) issued the New Entrant Program Interim Final Rule in response to the requirement in the Motor Carrier Safety Improvement Act of 1999. The requirement in the Act was based on research fi...

  4. Quantity and management of spent fuel from prototype and research reactors in Germany

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Dorr, Sabine; Bollingerfehr, Wilhelm; Filbert, Wolfgang

    Within the scope of an R and D project (project identification number FKZ 02 S 8679) sponsored by BMBF (Federal Ministry of Education and Research), the current state of storage and management of fuel elements from prototype and research reactors was established, and an approach for their future storage/management was developed. The spent fuels from prototype and research reactors in Germany that require disposal were specified and were described in regard to their repository-relevant characteristics. As there are currently no casks licensed for disposal in Germany, descriptions of casks that were considered to be suitable were provided. Based on themore » information provided on the spent fuel from prototype and research reactors and the potential casks, a technical disposal concept was developed. In this context, concepts to integrate the spent fuel from prototype and research reactors into existing disposal concepts for spent fuel from German nuclear power plants and for waste from reprocessing were developed for salt and clay formations. (authors)« less

  5. ORNL Named as Part of IAES Research Reactor Hub

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    None

    The International Atomic Energy Agency (IAEA) has named ORNL and Idaho National Laboratory part of an International Centre based on Research Reactors. The designation makes the United States one of only three countries identified for unique capabilities and excellence in nuclear research.

  6. Advantages of liquid fluoride thorium reactor in comparison with light water reactor

    NASA Astrophysics Data System (ADS)

    Bahri, Che Nor Aniza Che Zainul; Majid, Amran Ab.; Al-Areqi, Wadeeah M.

    2015-04-01

    Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

  7. Suggestion on the safety classification of spent fuel dry storage in China’s pressurized water reactor nuclear power plant

    NASA Astrophysics Data System (ADS)

    Liu, Ting; Qu, Yunhuan; Meng, De; Zhang, Qiaoer; Lu, Xinhua

    2018-01-01

    China’s spent fuel storage in the pressurized water reactors(PWR) is stored with wet storage way. With the rapid development of nuclear power industry, China’s NPPs(NPPs) will not be able to meet the problem of the production of spent fuel. Currently the world’s major nuclear power countries use dry storage as a way of spent fuel storage, so in recent years, China study on additional spent fuel dry storage system mainly. Part of the PWR NPP is ready to apply for additional spent fuel dry storage system. It also need to safety classificate to spent fuel dry storage facilities in PWR, but there is no standard for safety classification of spent fuel dry storage facilities in China. Because the storage facilities of the spent fuel dry storage are not part of the NPP, the classification standard of China’s NPPs is not applicable. This paper proposes the safety classification suggestion of the spent fuel dry storage for China’s PWR NPP, through to the study on China’s safety classification principles of PWR NPP in “Classification for the items of pressurized water reactor nuclear power plants (GB/T 17569-2013)”, and safety classification about spent fuel dry storage system in NUREG/CR - 6407 in the United States.

  8. Status of reduced enrichment programs for research reactors in Japan

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Kanda, Keiji; Nishihara, Hedeaki; Shirai, Eiji

    1997-08-01

    The reduced enrichment programs for the JRR-2, JRR-3, JRR-4 and JMTR of Japan Atomic Energy Research Institute (JAERI), and the KUR of Kyoto University Research Reactor Institute (KURRI) have been partially completed and are mostly still in progress under the Joint Study Programs with Argonne National Laboratory (ANL). The JMTR and JRR-2 have been already converted to use MEU aluminide fuels in 1986 and 1987, respectively. The operation of the upgraded JRR-3(JRR-3M) has started in March 1990 with the LEU aluminide fuels. Since May 1992, the two elements have been inserted in the KUR. The safety review application for themore » full core conversion to use LEU silicide in the JMTR was approved in February 1992 and the conversion has been done in January 1994. The Japanese Government approved a cancellation of the KUHFR Project in February 1991, and in April 1994 the U.S. Government gave an approval to utilize HEU in the KUR instead of the KUHFR. Therefore, the KUR will be operated with HEU fuel until 2001. Since March 1994, Kyoto University is continuing negotiation with UKAEA Dounreay on spent fuel reprocessing and blending down of recovered uranium, in addition to that with USDOE.« less

  9. Reactor Operations Monitoring System

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Hart, M.M.

    1989-01-01

    The Reactor Operations Monitoring System (ROMS) is a VME based, parallel processor data acquisition and safety action system designed by the Equipment Engineering Section and Reactor Engineering Department of the Savannah River Site. The ROMS will be analyzing over 8 million signal samples per minute. Sixty-eight microprocessors are used in the ROMS in order to achieve a real-time data analysis. The ROMS is composed of multiple computer subsystems. Four redundant computer subsystems monitor 600 temperatures with 2400 thermocouples. Two computer subsystems share the monitoring of 600 reactor coolant flows. Additional computer subsystems are dedicated to monitoring 400 signals from assortedmore » process sensors. Data from these computer subsystems are transferred to two redundant process display computer subsystems which present process information to reactor operators and to reactor control computers. The ROMS is also designed to carry out safety functions based on its analysis of process data. The safety functions include initiating a reactor scram (shutdown), the injection of neutron poison, and the loadshed of selected equipment. A complete development Reactor Operations Monitoring System has been built. It is located in the Program Development Center at the Savannah River Site and is currently being used by the Reactor Engineering Department in software development. The Equipment Engineering Section is designing and fabricating the process interface hardware. Upon proof of hardware and design concept, orders will be placed for the final five systems located in the three reactor areas, the reactor training simulator, and the hardware maintenance center.« less

  10. Autonomous Control of Space Nuclear Reactors

    NASA Technical Reports Server (NTRS)

    Merk, John

    2013-01-01

    safety requirements of a nuclear reactor and provides high availability to the host system. The RICS is intended to interface with a host computer (the computer of the spacecraft where the reactor is mounted). The RICS leverages the safety features inherent in Earth-based reactors and also integrates the wide range neutron detector (WRND). A neutron detector provides the input that allows the RICS to do its job. The RICS is based on proven technology currently in use at a nuclear research facility. In its most basic form, the RICS is a ruggedized, compact data-acquisition and control system that could be adapted to support a wide variety of harsh environments. As such, the RICS could be a useful instrument outside the scope of a nuclear reactor, including military applications where failsafe data acquisition and control is required with stringent size, weight, and power constraints.

  11. 78 FR 40743 - Mine Safety and Health Research Advisory Committee, National Institute for Occupational Safety...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-07-08

    ... DEPARTMENT OF HEALTH AND HUMAN SERVICES Centers for Disease Control and Prevention Mine Safety and Health Research Advisory Committee, National Institute for Occupational Safety and Health (MSHRAC, NIOSH... Director, NIOSH, on priorities in mine safety and health research, including grants and contracts for such...

  12. Compendium of Traffic Safety Research Projects 1985-2013.

    DOT National Transportation Integrated Search

    2014-01-01

    Through many name changes, from the Office of Program Development and Evaluation, the Office of Research and Evaluation, to the current, Office of Behavioral Safety Research, our focus has remained on improving the safety of drivers, occupants, pedes...

  13. Advantages of liquid fluoride thorium reactor in comparison with light water reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Bahri, Che Nor Aniza Che Zainul, E-mail: anizazainul@gmail.com; Majid, Amran Ab.; Al-Areqi, Wadeeah M.

    2015-04-29

    Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclearmore » waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.« less

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Smith, Cyrus M; Nanstad, Randy K; Clayton, Dwight A

    2012-09-01

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

  15. Advanced Reactor Technologies - Regulatory Technology Development Plan (RTDP)

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Moe, Wayne L.

    This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a “critical path” for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However,more » it is also important to remember that certain “minimum” levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial “first step” in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by technology development studies, the anticipated

  16. Advanced Reactor Technology -- Regulatory Technology Development Plan (RTDP)

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Moe, Wayne Leland

    This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a “critical path” for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However,more » it is also important to remember that certain “minimum” levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial “first step” in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by technology development studies, the anticipated

  17. Fifty years of driving safety research.

    PubMed

    Lee, John D

    2008-06-01

    This brief review covers the 50 years of driving-related research published in Human Factors, its contribution to driving safety, and emerging challenges. Many factors affect driving safety, making it difficult to assess the impact of specific factors such as driver age, cell phone distractions, or collision warnings. The author considers the research themes associated with the approximately 270 articles on driving published in Human Factors in the past 50 years. To a large extent, current and past research has explored similar themes and concepts. Many articles published in the first 25 years focused on issues such as driver impairment, individual differences, and perceptual limits. Articles published in the past 25 years address similar issues but also point toward vehicle technology that can exacerbate or mitigate the negative effect of these issues. Conceptual and computational models have played an important role in this research. Improved crash-worthiness has contributed to substantial improvements in driving safety over the past 50 years, but future improvements will depend on enhancing driver performance and perhaps, more important, improving driver behavior. Developing models to guide this research will become more challenging as new technology enters the vehicle and shifts the focus from driver performance to driver behavior. Over the past 50 years, Human Factors has accumulated a large base of driving-related research that remains relevant for many of today's design and policy concerns.

  18. On Study of Application of Micro-reactor in Chemistry and Chemical Field

    NASA Astrophysics Data System (ADS)

    Zhang, Yunshen

    2018-02-01

    Serving as a micro-scale chemical reaction system, micro-reactor is characterized by high heat transfer efficiency and mass transfer, strictly controlled reaction time and good safety performance; compared with the traditional mixing reactor, it can effectively shorten reaction time by virtue of these advantages and greatly enhance the chemical reaction conversion rate. However, problems still exist in the process where micro-reactor is used for production in chemistry and chemical field, and relevant researchers are required to optimize and perfect the performance of micro-reactor. This paper analyzes specific application of micro-reactor in chemistry and chemical field.

  19. Convective cooling in a pool-type research reactor

    NASA Astrophysics Data System (ADS)

    Sipaun, Susan; Usman, Shoaib

    2016-01-01

    A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U3Si2Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system's performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm-3. An MSTR model consisting of 20% of MSTR's nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s-1 from the 4" pipe, and predicted pool surface temperature not exceeding 30°C.

  20. Research and Development Roadmaps for Liquid Metal Cooled Fast Reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Kim, T. K.; Grandy, C.; Natesan, K.

    The United States Department of Energy (DOE) commissioned the development of technology roadmaps for advanced (non-light water reactor) reactor concepts to help focus research and development funding over the next five years. The roadmaps show the research and development needed to support demonstration of an advanced (non-LWR) concept by the early 2030s, consistent with DOE’s Vision and Strategy for the Development and Deployment of Advanced Reactors. The intent is only to convey the technical steps that would be required to achieve such a goal; the means by which DOE will determine whether to invest in specific tasks will be treatedmore » separately. The starting point for the roadmaps is the Technical Readiness Assessment performed as part of an Advanced Test and Demonstration Reactor study released in 2016. The roadmaps were developed based upon a review of technical reports and vendor literature summarizing the technical maturity of each concept and the outstanding research and development needs. Critical path tasks for specific systems were highlighted on the basis of time and resources needed to complete the tasks and the importance of the system to the performance of the reactor concept. The roadmaps are generic, i.e. not specific to a particular vendor’s design but vendor design information may have been used as representative of the concept family. In the event that both near-term and more advanced versions of a concept are being developed, either a single roadmap with multiple branches or separate roadmaps for each version were developed. In each case, roadmaps point to a demonstration reactor (engineering or commercial) and show the activities that must be completed in parallel to support that demonstration in the 2030-2035 window. This report provides the roadmaps for two fast reactor concepts, the Sodium-cooled Fast Reactor (SFR) and the Lead-cooled Fast Reactor (LFR). The SFR technology is mature enough for commercial demonstration by the early

  1. Research notes : safety at high-speed intersections.

    DOT National Transportation Integrated Search

    2010-04-01

    A 2010 study for ODOT by researchers at the Oregon State University School of Civil and Construction Engineering titled, Evaluating Safety and Operations of High-Speed Signalized Intersections, examined effective means for improving safety at isolate...

  2. Multipurpose epithermal neutron beam on new research station at MARIA research reactor in Swierk-Poland

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Gryzinski, M.A.; Maciak, M.

    MARIA reactor is an open-pool research reactor what gives the chance to install uranium fission converter on the periphery of the core. It could be installed far enough not to induce reactivity of the core but close enough to produce high flux of fast neutrons. Special design of the converter is now under construction. It is planned to set the research stand based on such uranium converter in the near future: in 2015 MARIA reactor infrastructure should be ready (preparation started in 2013), in 2016 the neutron beam starts and in 2017 opening the stand for material and biological researchmore » or for medical training concerning BNCT. Unused for many years, horizontal channel number H2 at MARIA research rector in Poland, is going to be prepared as a part of unique stand. The characteristics of the neutron beam will be significant advantage of the facility. High flux of neutrons at the level of 2x10{sup 9} cm{sup -2}s{sup -1} will be obtainable by uranium neutron converter located 90 cm far from the reactor core fuel elements (still inside reactor core basket between so called core reflectors). Due to reaction of core neutrons with converter U{sub 3}Si{sub 2} material it will produce high flux of fast neutrons. After conversion neutrons will be collimated and moderated in the channel by special set of filters and moderators. At the end of H2 channel i.e. at the entrance to the research room neutron energy will be in the epithermal energy range with neutron intensity at least at the level required for BNCT (2x10{sup 9} cm{sup -2}s{sup -1}). For other purposes density of the neutron flux could be smaller. The possibility to change type and amount of installed filters/moderators which enables getting different properties of the beam (neutron energy spectrum, neutron-gamma ratio and beam profile and shape) is taken into account. H2 channel is located in separate room which is adjacent to two other empty rooms under the preparation for research laboratories (200 m2

  3. Manned space flight nuclear system safety. Volume 3: Reactor system preliminary nuclear safety analysis. Part 2: Accident Model Document (AMD)

    NASA Technical Reports Server (NTRS)

    1972-01-01

    The Accident Model Document is one of three documents of the Preliminary Safety Analysis Report (PSAR) - Reactor System as applied to a Space Base Program. Potential terrestrial nuclear hazards involving the zirconium hydride reactor-Brayton power module are identified for all phases of the Space Base program. The accidents/events that give rise to the hazards are defined and abort sequence trees are developed to determine the sequence of events leading to the hazard and the associated probabilities of occurence. Source terms are calculated to determine the magnitude of the hazards. The above data is used in the mission accident analysis to determine the most probable and significant accidents/events in each mission phase. The only significant hazards during the prelaunch and launch ascent phases of the mission are those which arise form criticality accidents. Fission product inventories during this time period were found to be very low due to very limited low power acceptance testing.

  4. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    D.M. McEligot; K. G. Condie; G. E. McCreery

    2005-10-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generationmore » IV program.« less

  5. Lessons from Fukushima for Improving the Safety of Nuclear Reactors

    NASA Astrophysics Data System (ADS)

    Lyman, Edwin

    2012-02-01

    The March 2011 accident at the Fukushima Daiichi nuclear power plant has revealed serious vulnerabilities in the design, operation and regulation of nuclear power plants. While some aspects of the accident were plant- and site-specific, others have implications that are broadly applicable to the current generation of nuclear plants in operation around the world. Although many of the details of the accident progression and public health consequences are still unclear, there are a number of lessons that can already be drawn. The accident demonstrated the need at nuclear plants for robust, highly reliable backup power sources capable of functioning for many days in the event of a complete loss of primary off-site and on-site electrical power. It highlighted the importance of detailed planning for severe accident management that realistically evaluates the capabilities of personnel to carry out mitigation operations under extremely hazardous conditions. It showed how emergency plans rooted in the assumption that only one reactor at a multi-unit site would be likely to experience a crisis fail miserably in the event of an accident affecting multiple reactor units simultaneously. It revealed that alternate water injection following a severe accident could be needed for weeks or months, generating large volumes of contaminated water that must be contained. And it reinforced the grim lesson of Chernobyl: that a nuclear reactor accident could lead to widespread radioactive contamination with profound implications for public health, the economy and the environment. While many nations have re-examined their policies regarding nuclear power safety in the months following the accident, it remains to be seen to what extent the world will take the lessons of Fukushima seriously and make meaningful changes in time to avert another, and potentially even worse, nuclear catastrophe.

  6. New reactor cavity cooling system having passive safety features using novel shape for HTGRs and VHTRs

    DOE PAGES

    Takamatsu, Kuniyoshi; Hu, Rui

    2014-11-27

    A new, highly efficient reactor cavity cooling system (RCCS) with passive safety features without a requirement for electricity and mechanical drive is proposed for high temperature gas cooled reactors (HTGRs) and very high temperature reactors (VHTRs). The RCCS design consists of continuous closed regions; one is an ex-reactor pressure vessel (RPV) region and another is a cooling region having heat transfer area to ambient air assumed at 40 (°C). The RCCS uses a novel shape to efficiently remove the heat released from the RPV with radiation and natural convection. Employing the air as the working fluid and the ambient airmore » as the ultimate heat sink, the new RCCS design strongly reduces the possibility of losing the heat sink for decay heat removal. Therefore, HTGRs and VHTRs adopting the new RCCS design can avoid core melting due to overheating the fuels. The simulation results from a commercial CFD code, STAR-CCM+, show that the temperature distribution of the RCCS is within the temperature limits of the structures, such as the maximum operating temperature of the RPV, 713.15 (K) = 440 (°C), and the heat released from the RPV could be removed safely, even during a loss of coolant accident (LOCA). Finally, when the RCCS can remove 600 (kW) of the rated nominal state even during LOCA, the safety review for building the HTTR could confirm that the temperature distribution of the HTTR is within the temperature limits of the structures to secure structures and fuels after the shutdown because the large heat capacity of the graphite core can absorb heat from the fuel in a short period. Therefore, the capacity of the new RCCS design would be sufficient for decay heat removal.« less

  7. Convective cooling in a pool-type research reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Sipaun, Susan, E-mail: susan@nm.gov.my; Usman, Shoaib, E-mail: usmans@mst.edu

    2016-01-22

    A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U{sub 3}Si{sub 2}Al) in the form of rectangular plates. Gaps between the plates allow coolant to passmore » through and carry away heat. A study was carried out to map out heat flow as well as to predict the system’s performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm{sup −3}. An MSTR model consisting of 20% of MSTR’s nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s{sup −1} from the 4” pipe, and predicted pool surface temperature not exceeding 30°C.« less

  8. Current and prospective safety issues at the HFBR

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Tichler, P.R.

    The Brookhaven High Flux Beam Reactor (HFBR) was designed primarily to produce external neutron beams for experimental research. It is cooled, moderated and reflected by heavy water and uses MTR-ETR type fuel elements containing enriched uranium. The reactor power when operation began in 19965 was 40 MW, was raised to 60 MW in 1982 after a number of plant modifications, and operated at that level until 1989. Since that time safety questions have been raised which resulted in extended shutdowns and a reduction in operating power to 30 MW. This paper will discuss the principle safety issues, plans for theirmore » resolution and return to 60 MW operation. In addition, radiation embrittlement of the reactor vessel and thermal shield and its affect on the life of the facility will be briefly discussed.« less

  9. 77 FR 40622 - Mine Safety and Health Research Advisory Committee, National Institute for Occupational Safety...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-07-10

    ... DEPARTMENT OF HEALTH AND HUMAN SERVICES Centers for Disease Control and Prevention Mine Safety and Health Research Advisory Committee, National Institute for Occupational Safety and Health (MSHRAC, NIOSH..., oxygen supply partnership, safety culture, occupational health and safety management systems, preventing...

  10. Recent upgrades and new scientific infrastructure of MARIA research reactor, Otwock-Swierk, Poland

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    NONE

    The MARIA reactor is open-pool type, water and beryllium moderated. It has two independent primary cooling systems: fuel and pool cooling system. Each fuel assembly is cooled down separately in pressurized channels with individual performances characterization. The fuel assemblies consist of five layers of bent plates or six concentric tubes. Currently it is one of the most powerful research reactors in Europe with operation availability at least up to 2030. Its nominal thermal power is 30 MW. It is characterized by high neutron flux density: up to 3x10{sup 14} n cm{sup -2} s{sup -1} in case of thermal neutrons, andmore » up to 2x10{sup 13} n cm{sup -2} s{sup -1} in case of fast neutrons. The reactor is operated for ca. 4000 h per year. The reactor facility is equipped with fully equipped three hot cells with shielding up to 10{sup 15} Bq. Adjacent to the reactor facility, the radio-pharmaceutics plant (POLATOM) and Material Research Laboratory are located. They are equipped with a number of hot cells with instrumentation. The transport system of radioactive materials from reactor facility to Material Research Laboratory is available. During 2014 the MARIA reactor has been operated with three different types of fuel the same time: previous 36% enriched fuel, and two types of new LEU fuels. In the meantime, molybdenum irradiation programme has been developed. Maria is a multifunctional research tool, with a notable application in production of radioisotopes, radio-pharmaceutics manufacturing (ca. 600 TBq/y), {sup 99}Mo for medical scintigraphy (ca. 6000 TBq/y), neutron transmutation doping of silicon single crystals, wide scientific research based on neutron beams utilization. From the beginning MARIA reactor was intended for loop and fuel testing research activities. Currently it is used mostly as material testing and irradiation facility and for that reason it has wide experimental capabilities. There are eight horizontal irradiation channels from among whom six

  11. A Research Reactor Concept to Support NTP Development

    NASA Technical Reports Server (NTRS)

    Eades, Michael J.; Blue, T. E.; Gerrish, Harold P.; Hardin, Leroy A.

    2014-01-01

    In support of efforts for research into the design and development of man rated Nuclear Thermal Propulsion (NTP), the National Aeronautics and Space Administration (NASA), Marshall Space Flight Center (MSFC), is evaluating the potential for building a Nuclear Regulatory Commission (NRC) licensed NTP based research reactor (NTPRR). The proposed NTPRR would be licensed by NASA and operated jointly by NASA and university partners. The purpose of the NTPRR would be used to perform further research into the technologies and systems needed for a successful NTP project and promote nuclear training and education.

  12. Chemical Safety for Sustainability Research Action Plan 2012-2016

    EPA Pesticide Factsheets

    EPA’s Chemical Safety for Sustainability (CSS) research program presents the purpose, design and themes of the Agency’s CSS research efforts to ensure safety in the design, manufacture and use of existing and future chemicals

  13. NUCLEAR REACTOR CONTROL SYSTEM

    DOEpatents

    Epler, E.P.; Hanauer, S.H.; Oakes, L.C.

    1959-11-01

    A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.

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

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Barariu, G.; Giumanca, R.

    2006-07-01

    Pre-feasibility and feasibility studies were performed for decommissioning of the water-cooled water-moderated research reactor (WWER) located in Bucharest - Magurele, Romania. Using these studies as a starting point, the preferred safe management strategy for radioactive wastes produced by reactor decommissioning is outlined. The strategy must account for reactor decommissioning, as well as for the rehabilitation of the existing Radioactive Waste Treatment Plant and for the upgrade of the Radioactive Waste Disposal Facility at Baita-Bihor. Furthermore, the final rehabilitation of the laboratories and ecological reconstruction of the grounds need to be provided for, in accordance with national and international regulations. Inmore » accordance with IAEA recommendations at the time, the pre-feasibility study proposed three stages of decommissioning. However, since then new ideas have surfaced with regard to decommissioning. Thus, taking into account the current IAEA ideology, the feasibility study proposes that decommissioning of the WWER be done in one stage to an unrestricted clearance level of the reactor building in an Immediate Dismantling option. Different options and the corresponding derived preferred option for waste management are discussed taking into account safety measures, but also considering technical, logistical and economic factors. For this purpose, possible types of waste created during each decommissioning stage are reviewed. An approximate inventory of each type of radioactive waste is presented. The proposed waste management strategy is selected in accordance with the recommended international basic safety standards identified in the previous phase of the project. The existing Radioactive Waste Treatment Plant (RWTP) from the Horia Hulubei Institute for Nuclear Physics and Engineering (IFIN-HH), which has been in service with no significant upgrade since 1974, will need refurbishing due to deterioration, as well as upgrading in order to

  15. JEN-1 Reactor Control System; SISTEMA DE CONTROL DEL REACTOR JEN-1

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Cantillo, M.F.; Nuno, C.M.; Andreu, J.L.M.

    1963-01-01

    ABS>The JEN-1 3Mw power swimming pool reactor electrical control circuits are described. Start-up, power generation in the core, and shutdown are controlled by the reactor control system. This control system guarantees in each moment the safety conditions during reactor operation. Each circuit was represented by a scheme, complemented with a description of its function, components, and operation theory. Components described include: scram circuit; fission counter control circuit; servo control circuit; control circuit of safety sheets; control circuits of primary, secondary, and clean-up pump motors and tower fan motor; primary valve motor circuit; center cubicle alarm circuit; and process alarm circuit.more » (auth)« less

  16. Current safety practices in nano-research laboratories in China.

    PubMed

    Zhang, Can; Zhang, Jing; Wang, Guoyu

    2014-06-01

    China has become a key player in the global nanotechnology field, however, no surveys have specifically examined safety practices in the Chinese nano-laboratories in depth. This study reports results of a survey of 300 professionals who work in research laboratories that handle nanomaterials in China. We recruited participants at three major nano-research laboratories (which carry out research in diverse fields such as chemistry, material science, and biology) and the nano-chemistry session of the national meeting of the Chinese Chemical Society. Results show that almost all nano-research laboratories surveyed had general safety regulations, whereas less than one third of respondents reported having nanospecific safety rules. General safety measures were in place in most surveyed nano-research laboratories, while nanospecific protective measures existed or were implemented less frequently. Several factors reported from the scientific literature including nanotoxicology knowledge gaps, technical limitations on estimating nano-exposure, and the lack of nano-occupational safety legislation may contribute to the current state of affairs. With these factors in mind and embracing the precautionary principle, we suggest strengthening or providing nanosafety training (including raising risk awareness) and establishing nanosafety guidelines in China, to better protect personnel in the nano-workplace.

  17. U-Mo Monolithic Fuel for Nuclear Research and Test Reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Prabhakaran, Ramprashad

    The metallic fuel selected to replace the current HEU fuels in the research and test reactors is the LEU-10 weight % Mo alloy in the form of a thin sheet or foil encapsulated in AA6061 aluminum alloy with a zirconium interlayer. In order to effectively lead this pursuit, new developments in processing and fabrication of the fuel elements have been initiated, along with a better understanding of material behavior before and after irradiation as a result of these new developments. This editorial note gives an introduction about research and test reactors, need for HEU to LEU conversion, fuel requirements, highmore » uranium density monolithic fuel development and an overview of the four articles published in the December 2017 issue of JOM under a special topic titled “U-Mo Monolithic Fuel for Nuclear Research and Test Reactors”.« less

  18. Irradiation Tests Supporting LEU Conversion of Very High Power Research Reactors in the US

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Woolstenhulme, N. E.; Cole, J. I.; Glagolenko, I.

    The US fuel development team is developing a high density uranium-molybdenum alloy monolithic fuel to enable conversion of five high-power research reactors. Previous irradiation tests have demonstrated promising behavior for this fuel design. A series of future irradiation tests will enable selection of final fuel fabrication process and provide data to qualify the fuel at moderately-high power conditions for use in three of these five reactors. The remaining two reactors, namely the Advanced Test Reactor and High Flux Isotope Reactor, require additional irradiation tests to develop and demonstrate the fuel’s performance with even higher power conditions, complex design features, andmore » other unique conditions. This paper reviews the program’s current irradiation testing plans for these moderately-high irradiation conditions and presents conceptual testing strategies to illustrate how subsequent irradiation tests will build upon this initial data package to enable conversion of these two very-high power research reactors.« less

  19. Translating Health Services Research into Practice in the Safety Net.

    PubMed

    Moore, Susan L; Fischer, Ilana; Havranek, Edward P

    2016-02-01

    To summarize research relating to health services research translation in the safety net through analysis of the literature and case study of a safety net system. Literature review and key informant interviews at an integrated safety net hospital. This paper describes the results of a comprehensive literature review of translational science literature as applied to health care paired with qualitative analysis of five key informant interviews conducted with senior-level management at Denver Health and Hospital Authority. Results from the literature suggest that implementing innovation may be more difficult in the safety net due to multiple factors, including financial and organizational constraints. Results from key informant interviews confirmed the reality of financial barriers to innovation implementation but also implied that factors, including institutional respect for data, organizational attitudes, and leadership support, could compensate for disadvantages. Translating research into practice is of critical importance to safety net providers, which are under increased pressure to improve patient care and satisfaction. Results suggest that translational research done in the safety net can better illuminate the special challenges of this setting; more such research is needed. © Health Research and Educational Trust.

  20. Annotated Bibliography of Rail Transit Safety, 1975-1980, with Emphasis on Safety Research and Development.

    DOT National Transportation Integrated Search

    1981-09-01

    The bibliography provides a comprehensive review of published literature concerning rail transit safety and includes 186 annotated entries. The report covers domestic and foreign material on rail transit safety and related safety research and develop...

  1. Initial verification and validation of RAZORBACK - A research reactor transient analysis code

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Talley, Darren G.

    2015-09-01

    This report describes the work and results of the initial verification and validation (V&V) of the beta release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This initial V&V effort was intended to confirm that the code work to-date shows good agreement between simulation and actualmore » ACRR operations, indicating that the subsequent V&V effort for the official release of the code will be successful.« less

  2. Time series modeling in traffic safety research.

    PubMed

    Lavrenz, Steven M; Vlahogianni, Eleni I; Gkritza, Konstantina; Ke, Yue

    2018-08-01

    The use of statistical models for analyzing traffic safety (crash) data has been well-established. However, time series techniques have traditionally been underrepresented in the corresponding literature, due to challenges in data collection, along with a limited knowledge of proper methodology. In recent years, new types of high-resolution traffic safety data, especially in measuring driver behavior, have made time series modeling techniques an increasingly salient topic of study. Yet there remains a dearth of information to guide analysts in their use. This paper provides an overview of the state of the art in using time series models in traffic safety research, and discusses some of the fundamental techniques and considerations in classic time series modeling. It also presents ongoing and future opportunities for expanding the use of time series models, and explores newer modeling techniques, including computational intelligence models, which hold promise in effectively handling ever-larger data sets. The information contained herein is meant to guide safety researchers in understanding this broad area of transportation data analysis, and provide a framework for understanding safety trends that can influence policy-making. Copyright © 2017 Elsevier Ltd. All rights reserved.

  3. Informatics for patient safety: a nursing research perspective.

    PubMed

    Bakken, Suzanne

    2006-01-01

    In Crossing the Quality Chasm, the Institute of Medicine (IOM) Committee on Quality of Health Care in America identified the critical role of information technology in designing a health system that produces care that is "safe, effective, patient-centered, timely, efficient, and equitable" (Committee on Quality of Health Care in America, 2001, p. 164). A subsequent IOM report contends that improved information systems are essential to a new health care delivery system that "both prevents errors and learns from them when they occur" (Committee on Data Standards for Patient Safety, 2004, p. 1). This review specifically highlights the role of informatics processes and information technology in promoting patient safety and summarizes relevant nursing research. First, the components of an informatics infrastructure for patient safety are described within the context of the national framework for delivering consumer-centric and information-rich health care and using the National Health Information Infrastructure (NHII) (Thompson & Brailer, 2004). Second, relevant nursing research is summarized; this includes research studies that contributed to the development of selected infrastructure components as well as studies specifically focused on patient safety. Third, knowledge gaps and opportunities for nursing research are identified for each main topic. The health information technologies deployed as part of the national framework must support nursing practice in a manner that enables prevention of medical errors and promotion of patient safety and contributes to the development of practice-based nursing knowledge as well as best practices for patient safety. The seminal work that has been completed to date is necessary, but not sufficient, to achieve this objective.

  4. Safety features of subcritical fluid fueled systems

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Bell, C.R.

    1995-10-01

    Accelerator-driven transmutation technology has been under study at Los Alamos for several years for application to nuclear waste treatment, tritium production, energy generation, and recently, to the disposition of excess weapons plutonium. Studies and evaluations performed to date at Los Alamos have led to a current focus on a fluid-fuel, fission system operating in a neutron source-supported subcritical mode, using molten salt reactor technology and accelerator-driven proton-neutron spallation. In this paper, the safety features and characteristics of such systems are explored from the perspective of the fundamental nuclear safety objectives that any reactor-type system should address. This exploration is qualitativemore » in nature and uses current vintage solid-fueled reactors as a baseline for comparison. Based on the safety perspectives presented, such systems should be capable of meeting the fundamental nuclear safety objectives. In addition, they should be able to provide the safety robustness desired for advanced reactors. However, the manner in which safety objectives and robustness are achieved is very different from that associated with conventional reactors. Also, there are a number of safety design and operational challenges that will have to be addressed for the safety potential of such systems to be credible.« less

  5. 77 FR 42771 - License Renewal for the Dow Chemical TRIGA Research Reactor

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-07-20

    ... Chemical Company in Midland, MI and is a part of the Analytical Sciences Laboratory. The reactor is housed...-Radiological Impacts The Dow TRIGA Research Reactor core is cooled by a light water primary system consisting... provided by the volume of primary coolant allows several hours of full-power operation without any...

  6. Management of nanomaterials safety in research environment.

    PubMed

    Groso, Amela; Petri-Fink, Alke; Magrez, Arnaud; Riediker, Michael; Meyer, Thierry

    2010-12-10

    Despite numerous discussions, workshops, reviews and reports about responsible development of nanotechnology, information describing health and environmental risk of engineered nanoparticles or nanomaterials is severely lacking and thus insufficient for completing rigorous risk assessment on their use. However, since preliminary scientific evaluations indicate that there are reasonable suspicions that activities involving nanomaterials might have damaging effects on human health; the precautionary principle must be applied. Public and private institutions as well as industries have the duty to adopt preventive and protective measures proportionate to the risk intensity and the desired level of protection. In this work, we present a practical, 'user-friendly' procedure for a university-wide safety and health management of nanomaterials, developed as a multi-stakeholder effort (government, accident insurance, researchers and experts for occupational safety and health). The process starts using a schematic decision tree that allows classifying the nano laboratory into three hazard classes similar to a control banding approach (from Nano 3--highest hazard to Nano1--lowest hazard). Classifying laboratories into risk classes would require considering actual or potential exposure to the nanomaterial as well as statistical data on health effects of exposure. Due to the fact that these data (as well as exposure limits for each individual material) are not available, risk classes could not be determined. For each hazard level we then provide a list of required risk mitigation measures (technical, organizational and personal). The target 'users' of this safety and health methodology are researchers and safety officers. They can rapidly access the precautionary hazard class of their activities and the corresponding adequate safety and health measures. We succeed in convincing scientist dealing with nano-activities that adequate safety measures and management are promoting

  7. Management of nanomaterials safety in research environment

    PubMed Central

    2010-01-01

    Despite numerous discussions, workshops, reviews and reports about responsible development of nanotechnology, information describing health and environmental risk of engineered nanoparticles or nanomaterials is severely lacking and thus insufficient for completing rigorous risk assessment on their use. However, since preliminary scientific evaluations indicate that there are reasonable suspicions that activities involving nanomaterials might have damaging effects on human health; the precautionary principle must be applied. Public and private institutions as well as industries have the duty to adopt preventive and protective measures proportionate to the risk intensity and the desired level of protection. In this work, we present a practical, 'user-friendly' procedure for a university-wide safety and health management of nanomaterials, developed as a multi-stakeholder effort (government, accident insurance, researchers and experts for occupational safety and health). The process starts using a schematic decision tree that allows classifying the nano laboratory into three hazard classes similar to a control banding approach (from Nano 3 - highest hazard to Nano1 - lowest hazard). Classifying laboratories into risk classes would require considering actual or potential exposure to the nanomaterial as well as statistical data on health effects of exposure. Due to the fact that these data (as well as exposure limits for each individual material) are not available, risk classes could not be determined. For each hazard level we then provide a list of required risk mitigation measures (technical, organizational and personal). The target 'users' of this safety and health methodology are researchers and safety officers. They can rapidly access the precautionary hazard class of their activities and the corresponding adequate safety and health measures. We succeed in convincing scientist dealing with nano-activities that adequate safety measures and management are promoting

  8. Thermos reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Labrousse, M.; Lerouge, B.; Dupuy, G.

    1978-04-01

    THERMOS is a water reactor designed to provide hot water up to 120/sup 0/C for district heating or for desalination applications. It is a 100-MW reactor based on proven technology: oxide fuel plate elements, integrated primary circuit, and reactor vessel located in the bottom of a pool. As in swimming pool reactors, the pool is used for biological shielding, emergency core cooling, and fission product filtering (in case of an accident). Before economics, safety is the main characteristic of the concept: no fuel failure admitted, core under water in any accidental configuration, inspection of every ''nuclear'' component, and double-wall containment.

  9. Sociotechnical approaches to workplace safety: Research needs and opportunities.

    PubMed

    Robertson, Michelle M; Hettinger, Lawrence J; Waterson, Patrick E; Noy, Y Ian; Dainoff, Marvin J; Leveson, Nancy G; Carayon, Pascale; Courtney, Theodore K

    2015-01-01

    The sociotechnical systems perspective offers intriguing and potentially valuable insights into problems associated with workplace safety. While formal sociotechnical systems thinking originated in the 1950s, its application to the analysis and design of sustainable, safe working environments has not been fully developed. To that end, a Hopkinton Conference was organised to review and summarise the state of knowledge in the area and to identify research priorities. A group of 26 international experts produced collaborative articles for this special issue of Ergonomics, and each focused on examining a key conceptual, methodological and/or theoretical issue associated with sociotechnical systems and safety. In this concluding paper, we describe the major conference themes and recommendations. These are organised into six topic areas: (1) Concepts, definitions and frameworks, (2) defining research methodologies, (3) modelling and simulation, (4) communications and decision-making, (5) sociotechnical attributes of safe and unsafe systems and (6) potential future research directions for sociotechnical systems research. Sociotechnical complexity, a characteristic of many contemporary work environments, presents potential safety risks that traditional approaches to workplace safety may not adequately address. In this paper, we summarise the investigations of a group of international researchers into questions associated with the application of sociotechnical systems thinking to improve worker safety.

  10. Sociotechnical approaches to workplace safety: Research needs and opportunities

    PubMed Central

    Robertson, Michelle M.; Hettinger, Lawrence J.; Waterson, Patrick E.; Ian Noy, Y.; Dainoff, Marvin J.; Leveson, Nancy G.; Carayon, Pascale; Courtney, Theodore K.

    2015-01-01

    The sociotechnical systems perspective offers intriguing and potentially valuable insights into problems associated with workplace safety. While formal sociotechnical systems thinking originated in the 1950s, its application to the analysis and design of sustainable, safe working environments has not been fully developed. To that end, a Hopkinton Conference was organised to review and summarise the state of knowledge in the area and to identify research priorities. A group of 26 international experts produced collaborative articles for this special issue of Ergonomics, and each focused on examining a key conceptual, methodological and/or theoretical issue associated with sociotechnical systems and safety. In this concluding paper, we describe the major conference themes and recommendations. These are organised into six topic areas: (1) Concepts, definitions and frameworks, (2) defining research methodologies, (3) modelling and simulation, (4) communications and decision-making, (5) sociotechnical attributes of safe and unsafe systems and (6) potential future research directions for sociotechnical systems research. Practitioner Summary: Sociotechnical complexity, a characteristic of many contemporary work environments, presents potential safety risks that traditional approaches to workplace safety may not adequately address. In this paper, we summarise the investigations of a group of international researchers into questions associated with the application of sociotechnical systems thinking to improve worker safety. PMID:25728246

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

    DOEpatents

    Forsberg, C.W.

    1985-02-19

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

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

    DOEpatents

    Forsberg, Charles W.

    1987-01-01

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

  13. Overview of Experiments for Physics of Fast Reactors from the International Handbooks of Evaluated Criticality Safety Benchmark Experiments and Evaluated Reactor Physics Benchmark Experiments

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Bess, J. D.; Briggs, J. B.; Gulliford, J.

    mentioning is the critical experiments with fast reactor fuel rods in water, interesting in terms of justification of nuclear safety during transportation and storage of fresh and spent fuel. These reports provide a detailed review of the experiment, designate the area of their application and include results of calculations on modern systems of constants in comparison with the estimated experimental data.« less

  14. 75 FR 27368 - Aerotest Operations, Inc., Aerotest Radiography and Research Reactor; Notice of Consideration of...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-05-14

    ...., Aerotest Radiography and Research Reactor; Notice of Consideration of Approval of Transfer and Conforming Amendment, Opportunity for a Hearing, and Order Imposing Procedures for Access to Sensitive Unclassified Non... Manager, Research and Test Reactors Licensing Branch, Division of Policy and Rulemaking, Office of Nuclear...

  15. An Overview of INEL Fusion Safety R&D Facilities

    NASA Astrophysics Data System (ADS)

    McCarthy, K. A.; Smolik, G. R.; Anderl, R. A.; Carmack, W. J.; Longhurst, G. R.

    1997-06-01

    The Fusion Safety Program at the Idaho National Engineering Laboratory has the lead for fusion safety work in the United States. Over the years, we have developed several experimental facilities to provide data for fusion reactor safety analyses. We now have four major experimental facilities that provide data for use in safety assessments. The Steam-Reactivity Measurement System measures hydrogen generation rates and tritium mobilization rates in high-temperature (up to 1200°C) fusion relevant materials exposed to steam. The Volatilization of Activation Product Oxides Reactor Facility provides information on mobilization and transport and chemical reactivity of fusion relevant materials at high temperature (up to 1200°C) in an oxidizing environment (air or steam). The Fusion Aerosol Source Test Facility is a scaled-up version of VAPOR. The ion-implanta-tion/thermal-desorption system is dedicated to research into processes and phenomena associated with the interaction of hydrogen isotopes with fusion materials. In this paper we describe the capabilities of these facilities.

  16. Narrative review of the UK Patient Safety Research Portfolio.

    PubMed

    Waring, Justin; Rowley, Emma; Dingwall, Robert; Palmer, Cecily; Murcott, Toby

    2010-01-01

    The UK Patient Safety Research Portfolio (PSRP) commissioned 38 studies investigating the threats to patient safety in various clinical settings and evaluating safety-related service interventions. This paper reviews 27 of these studies, drawing out emergent and cross-cutting themes in terms of theory, research methods and thematic findings. Given the diversity of PSRP studies, the paper takes a narrative approach that allows for qualitative description, interpretation and synthesis of the studies and their findings. The theoretical review shows the majority of PSRP studies draw upon a patient safety 'orthodoxy', developed from the concepts and models associated with the human factors approach. The methodological review shows that a diverse range of research designs and techniques have been utilized. Although many follow in the 'scientific' tradition, interpretative, mixed and innovative methods have been integral to research. The thematic review of findings highlights significant contributions to knowledge in the areas of 'people', 'organizations', and 'technology'. As well as identifying the various sources of risk in the organization and delivery of patient care, the studies also evaluate and make recommendations about service change and improvement. The PSRP has provided the foundations for significant theoretical, methodological and empirical advances in the area of patient safety. The findings and recommendations make important contributions to policy formulation and implementation as well as professional and managerial practice. Through this body of research the PSRP has supported the formation and growth of a thriving research community across academic, policy and professional communities.

  17. Digital Signal Processing Methods for Safety Systems Employed in Nuclear Power Industry

    NASA Astrophysics Data System (ADS)

    Popescu, George

    Some of the major safety concerns in the nuclear power industry focus on the readiness of nuclear power plant safety systems to respond to an abnormal event, the security of special nuclear materials in used nuclear fuels, and the need for physical security to protect personnel and reactor safety systems from an act of terror. Routine maintenance and tests of all nuclear reactor safety systems are performed on a regular basis to confirm the ability of these systems to operate as expected. However, these tests do not determine the reliability of these safety systems and whether the systems will perform for the duration of an accident and whether they will perform their tasks without failure after being engaged. This research has investigated the progression of spindle asynchronous error motion determined from spindle accelerations to predict bearings failure onset. This method could be applied to coolant pumps that are essential components of emergency core cooling systems at all nuclear power plants. Recent security upgrades mandated by the Nuclear Regulatory Commission and the Department of Homeland Security have resulted in implementation of multiple physical security barriers around all of the commercial and research nuclear reactors in the United States. A second part of this research attempts to address an increased concern about illegal trafficking of Special Nuclear Materials (SNM). This research describes a multi element scintillation detector system designed for non - invasive (passive) gamma ray surveillance for concealed SNM that may be within an area or sealed in a package, vehicle or shipping container. Detection capabilities of the system were greatly enhanced through digital signal processing, which allows the combination of two very powerful techniques: 1) Compton Suppression (CS) and 2) Pulse Shape Discrimination (PSD) with less reliance on complicated analog instrumentation.

  18. FRA funded grade crossing safety & trespass prevention research (June 2007 - present).

    DOT National Transportation Integrated Search

    2015-05-01

    FRAs Office of R&D and Office of Railroad : Safety have been actively supporting highway-rail grade crossing safety and trespass : prevention research to improve safety. Below : is a list of technical reports and research results : from FRA-funded...

  19. Compendium of traffic safety research projects : 1987-1997

    DOT National Transportation Integrated Search

    1997-09-01

    The National Highway Traffic Safety Administration's (NHTSA) Research and Evaluation Division, Office of Research and Traffic Records, conducts research and evaluation projects dealing with human attitudes, behaviors, and failures (motor vehicle cras...

  20. Reduced enrichment for research and test reactors: Proceedings

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Not Available

    1993-07-01

    The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris{o} National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately.

  1. Initiating Event Analysis of a Lithium Fluoride Thorium Reactor

    NASA Astrophysics Data System (ADS)

    Geraci, Nicholas Charles

    The primary purpose of this study is to perform an Initiating Event Analysis for a Lithium Fluoride Thorium Reactor (LFTR) as the first step of a Probabilistic Safety Assessment (PSA). The major objective of the research is to compile a list of key initiating events capable of resulting in failure of safety systems and release of radioactive material from the LFTR. Due to the complex interactions between engineering design, component reliability and human reliability, probabilistic safety assessments are most useful when the scope is limited to a single reactor plant. Thus, this thesis will study the LFTR design proposed by Flibe Energy. An October 2015 Electric Power Research Institute report on the Flibe Energy LFTR asked "what-if?" questions of subject matter experts and compiled a list of key hazards with the most significant consequences to the safety or integrity of the LFTR. The potential exists for unforeseen hazards to pose additional risk for the LFTR, but the scope of this thesis is limited to evaluation of those key hazards already identified by Flibe Energy. These key hazards are the starting point for the Initiating Event Analysis performed in this thesis. Engineering evaluation and technical study of the plant using a literature review and comparison to reference technology revealed four hazards with high potential to cause reactor core damage. To determine the initiating events resulting in realization of these four hazards, reference was made to previous PSAs and existing NRC and EPRI initiating event lists. Finally, fault tree and event tree analyses were conducted, completing the logical classification of initiating events. Results are qualitative as opposed to quantitative due to the early stages of system design descriptions and lack of operating experience or data for the LFTR. In summary, this thesis analyzes initiating events using previous research and inductive and deductive reasoning through traditional risk management techniques to

  2. Oxidation of aluminum alloy cladding for research and test reactor fuel

    NASA Astrophysics Data System (ADS)

    Kim, Yeon Soo; Hofman, G. L.; Robinson, A. B.; Snelgrove, J. L.; Hanan, N.

    2008-08-01

    The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

  3. A research agenda on patient safety in primary care. Recommendations by the LINNEAUS collaboration on patient safety in primary care.

    PubMed

    Verstappen, Wim; Gaal, Sander; Bowie, Paul; Parker, Diane; Lainer, Miriam; Valderas, Jose M; Wensing, Michel; Esmail, Aneez

    2015-09-01

    Healthcare can cause avoidable serious harm to patients. Primary care is not an exception, and the relative lack of research in this area lends urgency to a better understanding of patient safety, the future research agenda and the development of primary care oriented safety programmes. To outline a research agenda for patient safety improvement in primary care in Europe and beyond. The LINNEAUS collaboration partners analysed existing research on epidemiology and classification of errors, diagnostic and medication errors, safety culture, and learning for and improving patient safety. We discussed ideas for future research in several meetings, workshops and congresses with LINNEAUS collaboration partners, practising GPs, researchers in this field, and policy makers. This paper summarizes and integrates the outcomes of the LINNEAUS collaboration on patient safety in primary care. It proposes a research agenda on improvement strategies for patient safety in primary care. In addition, it provides background information to help to connect research in this field with practicing GPs and other healthcare workers in primary care. Future research studies should target specific primary care domains, using prospective methods and innovative methods such as patient involvement.

  4. A research agenda on patient safety in primary care. Recommendations by the LINNEAUS collaboration on patient safety in primary care

    PubMed Central

    Verstappen, Wim; Gaal, Sander; Bowie, Paul; Parker, Diane; Lainer, Miriam; Valderas, Jose M.; Wensing, Michel; Esmail, Aneez

    2015-01-01

    ABSTRACT Background: Healthcare can cause avoidable serious harm to patients. Primary care is not an exception, and the relative lack of research in this area lends urgency to a better understanding of patient safety, the future research agenda and the development of primary care oriented safety programmes. Objective: To outline a research agenda for patient safety improvement in primary care in Europe and beyond. Methods: The LINNEAUS collaboration partners analysed existing research on epidemiology and classification of errors, diagnostic and medication errors, safety culture, and learning for and improving patient safety. We discussed ideas for future research in several meetings, workshops and congresses with LINNEAUS collaboration partners, practising GPs, researchers in this field, and policy makers. Results: This paper summarizes and integrates the outcomes of the LINNEAUS collaboration on patient safety in primary care. It proposes a research agenda on improvement strategies for patient safety in primary care. In addition, it provides background information to help to connect research in this field with practicing GPs and other healthcare workers in primary care. Conclusion: Future research studies should target specific primary care domains, using prospective methods and innovative methods such as patient involvement. PMID:26339841

  5. Research Reactor Preparations for the Air Shipment of Highly Enriched Uranium from Romania

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    K. J. Allen; I. Bolshinsky; L. L. Biro

    2010-03-01

    In June 2009 two air shipments transported both unirradiated (fresh) and irradiated (spent) Russian-origin highly enriched uranium (HEU) nuclear fuel from two research reactors in Romania to the Russian Federation for conversion to low enriched uranium. The Institute for Nuclear Research at Pitesti (SCN Pitesti) shipped 30.1 kg of HEU fresh fuel pellets to Dimitrovgrad, Russia and the Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH) shipped 23.7 kilograms of HEU spent fuel assemblies from the VVR S research reactor at Magurele, Romania, to Chelyabinsk, Russia. Both HEU shipments were coordinated by the Russian Research Reactor Fuel Returnmore » Program (RRRFR) as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), were managed in Romania by the National Commission for Nuclear Activities Control (CNCAN), and were conducted in cooperation with the Russian Federation State Corporation Rosatom and the International Atomic Energy Agency. Both shipments were transported by truck to and from respective commercial airports in Romania and the Russian Federation and stored at secure nuclear facilities in Russia until the material is converted into low enriched uranium. These shipments resulted in Romania becoming the 3rd country under the RRRFR program and the 14th country under the GTRI program to remove all HEU. This paper describes the research reactor preparations and license approvals that were necessary to safely and securely complete these air shipments of nuclear fuel.« less

  6. Generalized implementation of software safety policies

    NASA Technical Reports Server (NTRS)

    Knight, John C.; Wika, Kevin G.

    1994-01-01

    As part of a research program in the engineering of software for safety-critical systems, we are performing two case studies. The first case study, which is well underway, is a safety-critical medical application. The second, which is just starting, is a digital control system for a nuclear research reactor. Our goal is to use these case studies to permit us to obtain a better understanding of the issues facing developers of safety-critical systems, and to provide a vehicle for the assessment of research ideas. The case studies are not based on the analysis of existing software development by others. Instead, we are attempting to create software for new and novel systems in a process that ultimately will involve all phases of the software lifecycle. In this abstract, we summarize our results to date in a small part of this project, namely the determination and classification of policies related to software safety that must be enforced to ensure safe operation. We hypothesize that this classification will permit a general approach to the implementation of a policy enforcement mechanism.

  7. A Virtual Reality Framework to Optimize Design, Operation and Refueling of GEN-IV Reactors.

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Rizwan-uddin; Nick Karancevic; Stefano Markidis

    2008-04-23

    many GEN-IV candidate designs are currently under investigation. Technical issues related to material, safety and economics are being addressed at research laboratories, industry and in academia. After safety, economic feasibility is likely to be the most important crterion in the success of GEN-IV design(s). Lessons learned from the designers and operators of GEN-II (and GEN-III) reactors must play a vital role in achieving both safety and economic feasibility goals.

  8. How Can We Improve School Safety Research?

    ERIC Educational Resources Information Center

    Astor, Ron Avi; Guerra, Nancy; Van Acker, Richard

    2010-01-01

    The authors of this article consider how education researchers can improve school violence and school safety research by (a) examining gaps in theoretical, conceptual, and basic research on the phenomena of school violence; (b) reviewing key issues in the design and evaluation of evidence-based practices to prevent school violence; and (c)…

  9. Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Belles, Randy; Poore, III, Willis P.; Brown, Nicholas R.

    2017-03-01

    This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-basedmore » description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.« less

  10. New Reactor Physics Benchmark Data in the March 2012 Edition of the IRPhEP Handbook

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    John D. Bess; J. Blair Briggs; Jim Gulliford

    2012-11-01

    The International Reactor Physics Experiment Evaluation Project (IRPhEP) was established to preserve integral reactor physics experimental data, including separate or special effects data for nuclear energy and technology applications. Numerous experiments that have been performed worldwide, represent a large investment of infrastructure, expertise, and cost, and are valuable resources of data for present and future research. These valuable assets provide the basis for recording, development, and validation of methods. If the experimental data are lost, the high cost to repeat many of these measurements may be prohibitive. The purpose of the IRPhEP is to provide an extensively peer-reviewed set ofmore » reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. Contributors from around the world collaborate in the evaluation and review of selected benchmark experiments for inclusion in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook) [1]. Several new evaluations have been prepared for inclusion in the March 2012 edition of the IRPhEP Handbook.« less

  11. SP-100 program: Space reactor system and subsystem investigations

    NASA Astrophysics Data System (ADS)

    Harty, R. B.

    1983-09-01

    For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. The nuclear safety review/approval process that is required for a space reactor system is summarized. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that is expected and to provide information that could be usable in future programs.

  12. Establishing research priorities for patient safety in emergency medicine: a multidisciplinary consensus panel.

    PubMed

    Plint, Amy C; Stang, Antonia S; Calder, Lisa A

    2015-01-01

    Patient safety in the context of emergency medicine is a relatively new field of study. To date, no broad research agenda for patient safety in emergency medicine has been established. The objective of this study was to establish patient safety-related research priorities for emergency medicine. These priorities would provide a foundation for high-quality research, important direction to both researchers and health-care funders, and an essential step in improving health-care safety and patient outcomes in the high-risk emergency department (ED) setting. A four-phase consensus procedure with a multidisciplinary expert panel was organized to identify, assess, and agree on research priorities for patient safety in emergency medicine. The 19-member panel consisted of clinicians, administrators, and researchers from adult and pediatric emergency medicine, patient safety, pharmacy, and mental health; as well as representatives from patient safety organizations. In phase 1, we developed an initial list of potential research priorities by electronically surveying a purposeful and convenience sample of patient safety experts, ED clinicians, administrators, and researchers from across North America using contact lists from multiple organizations. We used simple content analysis to remove duplication and categorize the research priorities identified by survey respondents. Our expert panel reached consensus on a final list of research priorities through an in-person meeting (phase 3) and two rounds of a modified Delphi process (phases 2 and 4). After phases 1 and 2, 66 unique research priorities were identified for expert panel review. At the end of phase 4, consensus was reached for 15 research priorities. These priorities represent four themes: (1) methods to identify patient safety issues (five priorities), (2) understanding human and environmental factors related to patient safety (four priorities), (3) the patient perspective (one priority), and (4) interventions for

  13. Improving proliferation resistance of high breeding gain generation 4 reactors using blankets composed of light water reactor waste

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Hellesen, C.; Grape, S.; Haakanson, A.

    2013-07-01

    Fertile blankets can be used in fast reactors to enhance the breeding gain as well as the passive safety characteristics. However, such blankets typically result in the production of weapons grade plutonium. For this reason they are often excluded from Generation IV reactor designs. In this paper we demonstrate that using blankets manufactured directly from spent light water (LWR) reactor fuel it is possible to produce a plutonium product with non-proliferation characteristics on a par with spent LWR fuel of 30-50 MWd/kg burnup. The beneficial breeding and safety characteristics are retained. (authors)

  14. CERCA LEU fuel assemblies testing in Maria Reactor - safety analysis summary and testing program scope.

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Pytel, K.; Mieleszczenko, W.; Lechniak, J.

    2010-03-01

    The presented paper contains neutronic and thermal-hydraulic (for steady and unsteady states) calculation results prepared to support annex to Safety Analysis Report for MARIA reactor in order to obtain approval for program of testing low-enriched uranium (LEU) lead test fuel assemblies (LTFA) manufactured by CERCA. This includes presentation of the limits and operational constraints to be in effect during the fuel testing investigations. Also, the scope of testing program (which began in August 2009), including additional measurements and monitoring procedures, is described.

  15. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Montierth, Leland M.

    2016-07-19

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element designmore » for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.« less

  16. Safety Issues in Agricultural Education Laboratories: A Synthesis of Research.

    ERIC Educational Resources Information Center

    Dyer, James E.; Andreasen, Randall J.

    1999-01-01

    Synthesis of research on safety in agricultural education laboratories found most research focused on agricultural mechanics. Labs appeared to be potentially hazardous places, and teachers have inadequate knowledge of safety laws and ways to provide a safe working environment. (SK)

  17. Long lifetime fast spectrum reactor for lunar surface power system

    NASA Astrophysics Data System (ADS)

    Kambe, Mitsuru

    1993-01-01

    In the framework of innovative reactor research activities, a conceptual design study of fast spectrum reactor and primary system for 800 kWe lunar surface power system to be combined with potassium Rankine cycle power conversion has been conducted to meet the power requirements of the lunar base activities in the next century. The reactor subsystem is characterized by RAPID (Refueling by All Pins Integrated Design) concept to enhance inherent safety and to enable quick and simplifed refueling in every 10 years. RAPID concept affords power plant design lifetime of up to 30 years. Integrity of the reactor structure and replacement of failed primary circuits are also discussed. Substantial reduction in per-kWh cost on considering launch, emplacement, and final disposition can be expected by a long system lifetime.

  18. Methodological and Ethical Issues in Pediatric Medication Safety Research.

    PubMed

    Carpenter, Delesha; Gonzalez, Daniel; Retsch-Bogart, George; Sleath, Betsy; Wilfond, Benjamin

    2017-09-01

    In May 2016, the Eshelman School of Pharmacy at The University of North Carolina at Chapel Hill convened the PharmSci conference to address the topic of "methodological and ethical issues in pediatric medication safety research." A multidisciplinary group of experts representing a diverse array of perspectives, including those of the US Food and Drug Administration, children's hospitals, and academia, identified important considerations for pediatric medication safety research and opportunities to advance the field. This executive summary describes current challenges that clinicians and researchers encounter related to pediatric medication safety research and identifies innovative and ethically sound methodologies to address these challenges to improve children's health. This article addresses 5 areas: (1) pediatric drug development and drug trials; (2) conducting comparative effectiveness research in pediatric populations; (3) child and parent engagement on study teams; (4) improving communication with children and parents; and (5) assessing child-reported outcomes and adverse drug events. Copyright © 2017 by the American Academy of Pediatrics.

  19. Researching safety culture: deliberative dialogue with a restorative lens.

    PubMed

    Lorenzini, Elisiane; Oelke, Nelly D; Marck, Patricia Beryl; Dall'agnol, Clarice Maria

    2017-10-01

    Safety culture is a key component of patient safety. Many patient safety strategies in health care have been adapted from high-reliability organizations (HRO) such as aviation. However, to date, attempts to transform the cultures of health care settings through HRO approaches have had mixed results. We propose a methodological approach for safety culture research, which integrates the theory and practice of restoration science with the principles and methods of deliberative dialogue to support active engagement in critical reflection and collective debate. Our aim is to describe how these two innovative approaches in health services research can be used together to provide a comprehensive effective method to study and implement change in safety culture. Restorative research in health care integrates socio-ecological theory of complex adaptive systems concepts with collaborative, place-sensitive study of local practice contexts. Deliberative dialogue brings together all stakeholders to collectively develop solutions on an issue to facilitate change. Together these approaches can be used to actively engage people in the study of safety culture to gain a better understanding of its elements. More importantly, we argue that the synergistic use of these approaches offers enhanced potential to move health care professionals towards actionable strategies to improve patient safety within today's complex health care systems. © The Author 2017. Published by Oxford University Press in association with the International Society for Quality in Health Care. All rights reserved. For permissions, please e-mail: journals.permissions@oup.com.

  20. Assessment of nuclear reactor concepts for low power space applications

    NASA Technical Reports Server (NTRS)

    Klein, Andrew C.; Gedeon, Stephen R.; Morey, Dennis C.

    1988-01-01

    The results of a preliminary small reactor concepts feasibility and safety evaluation designed to provide a first order validation of the nuclear feasibility and safety of six small reactor concepts are given. These small reactor concepts have potential space applications for missions in the 1 to 20 kWe power output range. It was concluded that low power concepts are available from the U.S. nuclear industry that have the potential for meeting both the operational and launch safety space mission requirements. However, each design has its uncertainties, and further work is required. The reactor concepts must be mated to a power conversion technology that can offer safe and reliable operation.

  1. Nuclear Reactors. Revised.

    ERIC Educational Resources Information Center

    Hogerton, John F.

    This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: How Reactors Work; Reactor Design; Research, Teaching, and Materials Testing; Reactors (Research, Teaching and Materials); Production Reactors; Reactors for Electric Power…

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

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.; Hui, Marvin M.; Berglund, Robert C.

    1991-01-01

    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.

  3. SP-100 Program: space reactor system and subsystem investigations

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Harty, R.B.

    1983-09-30

    For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. This report summarizes the nuclear safety review/approval process that will be required for a space reactor system. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that can be expected and to provide information that could be usable in future programs.

  4. THE COOLING REQUIREMENTS AND PROCESS SYSTEMS OF THE SOUTH AFRICAN RESEARCH REACTOR, SAFARI 1

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Colley, J.R.

    1962-12-01

    The SAFARI 1 research reactor is cooled and moderated by light water. There are three process systems, a primary water system which cools the reactor core and surroundings, a pool water system, and a secondary water system which removes the heat from the primary and pool systems. The cooling requirements for the reactor core and experimental facilities are outlined, and the cooling and purification functions of the three process systems are described. (auth)

  5. Healthcare Databases for Drug Safety Research: Data Validity Assessment Remains Crucial.

    PubMed

    Rawson, Nigel S B; D'Arcy, Carl

    2018-04-30

    Administrative healthcare utilization databases are frequently used either individually or as a component of aggregated data for evaluating drug safety issues without taking into account their known deficiencies. All too often insufficient evidence is provided about their validity for the purposes for which they are used. The assessment of data validity is a key constituent that should be included in drug safety research studies and should take a broad multifaceted approach that encompasses both diagnostic and drug exposure data. Drug safety researchers need to continue advancing their knowledge of the data resources they use and to ensure that they and the users of their research understand the limitations of the data that are the foundation on which their research is built. Fundamental issues regarding data validity should be addressed in each use of administrative data for drug safety research.

  6. An investigation of reactivity effect due to inadvertent filling of the irradiation channels with water in NIRR-1 Nigeria Research Reactor-1.

    PubMed

    Iliyasu, U; Ibrahim, Y V; Umar, Sadiq; Agbo, S A; Jibrin, Y

    2017-05-01

    Investigation of reactivity variation due to flooding of the irradiation channels of Nigeria Research Reactor (NIRR-1) a low power miniature neutron source reactor (MNSR) located at the Centre for Energy Research and Training, Ahmadu Bello University, Zaria Nigeria using the MCNP code for High Enrich Uranium (HEU) and Low Enrich Uranium (LEU) core has been simulated in this present study. In this work, the excess reactivity worth of flooding HEU core for 1 inner, 2 inner, 3 inner, 4 inner and all inner are 0.318mk, 0.577mk, 0.318mk, 1.204mk and 1.503mk respectively, and outer irradiation channels are 0.119mk, 0.169mk, 0.348mk, 0.438mk and 0.418mk respectively, the highest excess reactivity result from flooding both inner and outer irradiation channels is 2.04mk (±1.72×10 -7 ), the excess reactivity for LEU core was 0.299mk, 0.568mk, 0.896mk, 1.195mk and 1.524mk in the inner irradiation channels, and the outer irradiation channels are 0.129mk, 0.189mk, 0.219mk, 0.269mk and 0.548mk where the highest excess reactivity was 1.942mk (±1.64×10 -7 ) resulting from flooding inner and outer irradiation channels. The reactivity induced by flooding of the irradiation channels of NIRR-1 with water is within design safety limit enshrined in Safety Analysis Report of NIRR-1. The results also compare well with literature. Copyright © 2017 Elsevier Ltd. All rights reserved.

  7. The near boiling reactor: Conceptual design of a small inherently safe nuclear reactor to extend the operational capability of the Victoria Class submarine

    NASA Astrophysics Data System (ADS)

    Cole, Christopher J. P.

    Nuclear power has several unique advantages over other air independent energy sources for nuclear combat submarines. An inherently safe, small nuclear reactor, capable of supply the hotel load of the Victoria Class submarines, has been conceptually developed. The reactor is designed to complement the existing diesel electric power generation plant presently onboard the submarine. The reactor, rated at greater than 1 MW thermal, will supply electricity to the submarine's batteries through an organic Rankine cycle energy conversion plant at 200 kW. This load will increase the operational envelope of the submarine by providing up to 28 continuous days submerged, allowing for an enhanced indiscretion ratio (ratio of time spent on the surface versus time submerged) and a limited under ice capability. The power plant can be fitted into the existing submarine by inserting a 6 m hull plug. With its simplistic design and inherent safety features, the reactor plant will require a minimal addition to the crew. The reactor employs TRISO fuel particles for increased safety. The light water coolant remains at atmospheric pressure, exiting the core at 96°C. Burn-up control and limiting excess reactivity is achieved through movable reflector plates. Shut down and regulatory control is achieved through the thirteen hafnium control rods. Inherent safety is achieved through the negative prompt and delayed temperature coefficients, as well as the negative void coefficient. During a transient, the boiling of the moderator results in a sudden drop in reactivity, essentially shutting down the reactor. It is this characteristic after which the reactor has been named. The design of the reactor was achieved through modelling using computer codes such as MCNP5, WIMS-AECL, FEMLAB, and MicroShield5, in addition to specially written software for kinetics, heat transfer and fission product poisoning calculations. The work has covered a broad area of research and has highlighted additional areas

  8. Temperature actuated automatic safety rod release

    DOEpatents

    Hutter, E.; Pardini, J.A.; Walker, D.E.

    1984-03-13

    A temperature-actuated apparatus is disclosed for releasably supporting a safety rod in a nuclear reactor, comprising a safety rod upper adapter having a retention means, a drive shaft which houses the upper adapter, and a bimetallic means supported within the drive shaft and having at least one ledge which engages a retention means of the safety rod upper adapter. A pre-determined increase in temperature causes the bimetallic means to deform so that the ledge disengages from the retention means, whereby the bimetallic means releases the safety rod into the core of the reactor.

  9. Temperature actuated automatic safety rod release

    DOEpatents

    Hutter, Ernest; Pardini, John A.; Walker, David E.

    1987-01-01

    A temperature-actuated apparatus is disclosed for releasably supporting a safety rod in a nuclear reactor, comprising a safety rod upper adapter having a retention means, a drive shaft which houses the upper adapter, and a bimetallic means supported within the drive shaft and having at least one ledge which engages a retention means of the safety rod upper adapter. A pre-determined increase in temperature causes the bimetallic means to deform so that the ledge disengages from the retention means, whereby the bimetallic means releases the safety rod into the core of the reactor.

  10. Nuclear safety for the space exploration initiative

    NASA Technical Reports Server (NTRS)

    Dix, Terry E.

    1991-01-01

    The results of a study to identify potential hazards arising from nuclear reactor power systems for use on the lunar and Martian surfaces, related safety issues, and resolutions of such issues by system design changes, operating procedures, and other means are presented. All safety aspects of nuclear reactor power systems from prelaunch ground handling to eventual disposal were examined consistent with the level of detail for SP-100 reactor design at the 1988 System Design Review and for launch vehicle and space transport vehicle designs and mission descriptions as defined in the 90-day Space Exploration Initiative (SEI) study. Information from previous aerospace nuclear safety studies was used where appropriate. Safety requirements for the SP-100 space nuclear reactor system were compiled. Mission profiles were defined with emphasis on activities after low earth orbit insertion. Accident scenarios were then qualitatively defined for each mission phase. Safety issues were identified for all mission phases with the aid of simplified event trees. Safety issue resolution approaches of the SP-100 program were compiled. Resolution approaches for those safety issues not covered by the SP-100 program were identified. Additionally, the resolution approaches of the SP-100 program were examined in light of the moon and Mars missions.

  11. The EUCLID/V1 Integrated Code for Safety Assessment of Liquid Metal Cooled Fast Reactors. Part 1: Basic Models

    NASA Astrophysics Data System (ADS)

    Mosunova, N. A.

    2018-05-01

    The article describes the basic models included in the EUCLID/V1 integrated code intended for safety analysis of liquid metal (sodium, lead, and lead-bismuth) cooled fast reactors using fuel rods with a gas gap and pellet dioxide, mixed oxide or nitride uranium-plutonium fuel under normal operation, under anticipated operational occurrences and accident conditions by carrying out interconnected thermal-hydraulic, neutronics, and thermal-mechanical calculations. Information about the Russian and foreign analogs of the EUCLID/V1 integrated code is given. Modeled objects, equation systems in differential form solved in each module of the EUCLID/V1 integrated code (the thermal-hydraulic, neutronics, fuel rod analysis module, and the burnup and decay heat calculation modules), the main calculated quantities, and also the limitations on application of the code are presented. The article also gives data on the scope of functions performed by the integrated code's thermal-hydraulic module, using which it is possible to describe both one- and twophase processes occurring in the coolant. It is shown that, owing to the availability of the fuel rod analysis module in the integrated code, it becomes possible to estimate the performance of fuel rods in different regimes of the reactor operation. It is also shown that the models implemented in the code for calculating neutron-physical processes make it possible to take into account the neutron field distribution over the fuel assembly cross section as well as other features important for the safety assessment of fast reactors.

  12. Mapping a Research Agenda for Home Care Safety: Perspectives from Researchers, Providers, and Decision Makers

    ERIC Educational Resources Information Center

    Macdonald, Marilyn; Lang, Ariella; MacDonald, Jo-Anne

    2011-01-01

    The purpose of this qualitative interpretive design was to explore the perspectives of researchers, health care providers, policy makers, and decision makers on key risks, concerns, and emerging issues related to home care safety that would inform a line of research inquiry. Defining safety specifically in this home care context has yet to be…

  13. Construction safety research in the United States: targeting the Hispanic workforce

    PubMed Central

    Brunette, M

    2004-01-01

    While it is known that Hispanics have a continuous growing participation in the construction workforce and that their fatal and non-fatal occupational injuries are higher than any other ethnic group, very little construction safety and health research has been conducted in the United States. Research that focuses on safety and health of Hispanic workers employed in the construction industry might prove beneficial in reducing injuries and promoting safe and decent workplaces for all. The purpose of this article was twofold. First, to propose a research agenda where topics such as surveillance, intervention research on high risk occupations, intervention effectiveness evaluation, design and development of effective and appropriate safety training and educational materials, and the socioeconomic impact of injuries and illnesses, are investigated among the Hispanic construction workforce. Second, to present relevant aspects inherent to this particular population that need to be incorporated into the design and development stages of any safety and health research initiative. They include the occupational, social, economic, and cultural background of Hispanic workers; use of a participatory approach, proper selection and use of translation methods; and conducting collaborative research. Certain limitations and challenges related to the availability of resources for conducting safety and health research on Hispanic workers are further discussed. PMID:15314054

  14. Construction safety research in the United States: targeting the Hispanic workforce.

    PubMed

    Brunette, M J

    2004-08-01

    While it is known that Hispanics have a continuous growing participation in the construction workforce and that their fatal and non-fatal occupational injuries are higher than any other ethnic group, very little construction safety and health research has been conducted in the United States. Research that focuses on safety and health of Hispanic workers employed in the construction industry might prove beneficial in reducing injuries and promoting safe and decent workplaces for all. The purpose of this article was twofold. First, to propose a research agenda where topics such as surveillance, intervention research on high risk occupations, intervention effectiveness evaluation, design and development of effective and appropriate safety training and educational materials, and the socioeconomic impact of injuries and illnesses, are investigated among the Hispanic construction workforce. Second, to present relevant aspects inherent to this particular population that need to be incorporated into the design and development stages of any safety and health research initiative. They include the occupational, social, economic, and cultural background of Hispanic workers; use of a participatory approach, proper selection and use of translation methods; and conducting collaborative research. Certain limitations and challenges related to the availability of resources for conducting safety and health research on Hispanic workers are further discussed.

  15. Nuclear Safety. Technical progress journal, April--June 1996: Volume 37, No. 2

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Muhlheim, M D

    1996-01-01

    This journal covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities.

  16. Conceptual design study of Fusion Experimental Reactor (FY86 FER): Safety

    NASA Astrophysics Data System (ADS)

    Seki, Yasushi; Iida, Hiromasa; Honda, Tsutomu

    1987-08-01

    This report describes the study on safety for FER (Fusion Experimental Reactor) which has been designed as a next step machine to the JT-60. Though the final purpose of this study is to have an image of design base accident, maximum credible accident and to assess their risk or probability, etc., as FER plant system, the emphasis of this years study is placed on fuel-gas circulation system where the tritium inventory is maximum. The report consists of two chapters. The first chapter summarizes the FER system and describes FMEA (Failure Mode and Effect Analysis) and related accident progression sequence for FER plant system as a whole. The second chapter of this report is focused on fuel-gas circulation system including purification, isotope separation and storage. Probability of risk is assessed by the probabilistic risk analysis (PRA) procedure based on FMEA, ETA and FTA.

  17. Propellant actuated nuclear reactor steam depressurization valve

    DOEpatents

    Ehrke, Alan C.; Knepp, John B.; Skoda, George I.

    1992-01-01

    A nuclear fission reactor combined with a propellant actuated depressurization and/or water injection valve is disclosed. The depressurization valve releases pressure from a water cooled, steam producing nuclear reactor when required to insure the safety of the reactor. Depressurization of the reactor pressure vessel enables gravity feeding of supplementary coolant water through the water injection valve to the reactor pressure vessel to prevent damage to the fuel core.

  18. IAEA coordinated research project on thermal-hydraulics of Supercritical Water-Cooled Reactors (SCWRs)

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Yamada, K.; Aksan, S. N.

    The Supercritical Water-Cooled Reactor (SCWR) is an innovative water-cooled reactor concept, which uses supercritical pressure water as reactor coolant. It has been attracting interest of many researchers in various countries mainly due to its benefits of high thermal efficiency and simple primary systems, resulting in low capital cost. The IAEA started in 2008 a Coordinated Research Project (CRP) on Thermal-Hydraulics of SCWRs as a forum to foster the exchange of technical information and international collaboration in research and development. This paper summarizes the activities and current status of the CRP, as well as major progress achieved to date. At present,more » 15 institutions closely collaborate in several tasks. Some organizations have been conducting thermal-hydraulics experiments and analysing the data, and others have been participating in code-to-test and/or code-to-code benchmark exercises. The expected outputs of the CRP are also discussed. Finally, the paper introduces several IAEA activities relating to or arising from the CRP. (authors)« less

  19. Design, Construction and Testing of an In-Pile Loop for PWR (Pressurized Water Reactor) Simulation.

    DTIC Science & Technology

    1987-06-01

    computer modeling remains at best semiempirical (C-i), this large variation in scaling factor makes extrapolation of data impossible. The DIDO Water...in a full scale PWR are not practical. The reactor plant is not controlled to tolerances necessary for research, and utilities are reluctant to vary...MIT Reactor Safeguards Committee, in revision 1 to the PCCL Safety Evaluation Report (SER), for final approval to begin in-pile testing and

  20. Human Factors Research in Anesthesia Patient Safety

    PubMed Central

    Weinger, Matthew B.; Slagle, Jason

    2002-01-01

    Patient safety has become a major public concern. Human factors research in other high-risk fields has demonstrated how rigorous study of factors that affect job performance can lead to improved outcome and reduced errors after evidence-based redesign of tasks or systems. These techniques have increasingly been applied to the anesthesia work environment. This paper describes data obtained recently using task analysis and workload assessment during actual patient care and the use of cognitive task analysis to study clinical decision making. A novel concept of “non-routine events” is introduced and pilot data are presented. The results support the assertion that human factors research can make important contributions to patient safety. Information technologies play a key role in these efforts.

  1. Human factors research in anesthesia patient safety.

    PubMed Central

    Weinger, M. B.; Slagle, J.

    2001-01-01

    Patient safety has become a major public concern. Human factors research in other high-risk fields has demonstrated how rigorous study of factors that affect job performance can lead to improved outcome and reduced errors after evidence-based redesign of tasks or systems. These techniques have increasingly been applied to the anesthesia work environment. This paper describes data obtained recently using task analysis and workload assessment during actual patient care and the use of cognitive task analysis to study clinical decision making. A novel concept of "non-routine events" is introduced and pilot data are presented. The results support the assertion that human factors research can make important contributions to patient safety. Information technologies play a key role in these efforts. PMID:11825287

  2. Proceedings of Twenty-Seventh Annual Institute on Mining Health, Safety and Research

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Bockosh, G.R.; Langton, J.; Karmis, M.

    1996-12-31

    This Proceedings contains the presentations made during the program of the Twenty-Seventh Annual Institute on Mining Health, Safety and Research held at Virginia Polytechnic Institute and State University, Blacksburg, Virginia, on August 26-28, 1996. The Twenty-Seventh Annual Institute on Mining, Health, Safety and Research was the latest in a series of conferences held at Virginia Polytechnic Institute and State University, cosponsored by the Mine Safety and Health Administration, United States Department of Labor, and the Pittsburgh Research Center, United States Department of Energy (formerly part of the Bureau of Mines, U. S. Department of Interior). The Institute provides an informationmore » forum for mine operators, managers, superintendents, safety directors, engineers, inspectors, researchers, teachers, state agency officials, and others with a responsible interest in the important field of mining health, safety and research. In particular, the Institute is designed to help mine operating personnel gain a broader knowledge and understanding of the various aspects of mining health and safety, and to present them with methods of control and solutions developed through research. Selected papers have been processed separately for inclusion in the Energy Science and Technology database.« less

  3. NHTSA's behavioral safety research: updated, annotated bibliography, 1985-2013 : traffic tech.

    DOT National Transportation Integrated Search

    2014-01-01

    Through many name changes, from the Office of Program : Development and Evaluation, the Office of Research and : Evaluation, to the current Office of Behavioral Safety Research, : our focus has remained on improving the safety of drivers, : occupants...

  4. TU-EF-BRD-04: Summing It Up: The Future of Quality and Safety Research

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Ford, E.

    Research related to quality and safety has been a staple of medical physics academic activities for a long time. From very early on, medical physicists have developed new radiation measurement equipment and analysis techniques, created ever increasingly accurate dose calculation models, and have vastly improved imaging, planning, and delivery techniques. These and other areas of interest have improved the quality and safety of radiotherapy for our patients. With the advent of TG-100, quality and safety is an area that will garner even more research interest in the future. As medical physicists pursue quality and safety research in greater numbers, itmore » is worthwhile to consider what actually constitutes research on quality and safety. For example, should the development of algorithms for real-time EPID-based in-vivo dosimetry be defined as “quality and safety” research? How about the clinical implementation of such as system? Surely the application of failure modes and effects analysis to a clinical process would be considered quality and safety research, but is this type of research that should be included in the medical physics peer-reviewed literature? The answers to such questions are of critical importance to set researchers in a direction that will provide the greatest benefit to our field and the patients we serve. The purpose of this symposium is to consider what constitutes research in the arena of quality and safety and differentiate it from other research directions. The key distinction here is developing the tool itself (e.g. algorithms for EPID dosimetry) vs. studying the impact of the tool with some quantitative metric. Only the latter would I call quality and safety research. Issues of ‘basic’ versus ‘applied’ quality and safety research will be covered as well as how the research results should be structured to provide increasing levels of support that a quality and safety intervention is effective and sustainable. Examples from

  5. TU-EF-BRD-01: Topics in Quality and Safety Research and Level of Evidence

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Pawlicki, T.

    Research related to quality and safety has been a staple of medical physics academic activities for a long time. From very early on, medical physicists have developed new radiation measurement equipment and analysis techniques, created ever increasingly accurate dose calculation models, and have vastly improved imaging, planning, and delivery techniques. These and other areas of interest have improved the quality and safety of radiotherapy for our patients. With the advent of TG-100, quality and safety is an area that will garner even more research interest in the future. As medical physicists pursue quality and safety research in greater numbers, itmore » is worthwhile to consider what actually constitutes research on quality and safety. For example, should the development of algorithms for real-time EPID-based in-vivo dosimetry be defined as “quality and safety” research? How about the clinical implementation of such as system? Surely the application of failure modes and effects analysis to a clinical process would be considered quality and safety research, but is this type of research that should be included in the medical physics peer-reviewed literature? The answers to such questions are of critical importance to set researchers in a direction that will provide the greatest benefit to our field and the patients we serve. The purpose of this symposium is to consider what constitutes research in the arena of quality and safety and differentiate it from other research directions. The key distinction here is developing the tool itself (e.g. algorithms for EPID dosimetry) vs. studying the impact of the tool with some quantitative metric. Only the latter would I call quality and safety research. Issues of ‘basic’ versus ‘applied’ quality and safety research will be covered as well as how the research results should be structured to provide increasing levels of support that a quality and safety intervention is effective and sustainable. Examples from

  6. Neutron transmutation doping of silicon in the SAFARI-1 research reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Louw, P.A.; Robertson, D.G.; Strydom, W.J.

    1994-12-31

    The SAFARI-1 research reactor has operated with an exemplary safety record since commissioning in 1965. As part of a commercialisation effort a silicon irradiation facility (SILIRAD) has been installed in the poolside region of SAFARI-1 for Neutron Transmutation Doping (NTD) of silicon. Commissioning of the facility took place in the last quarter of 1992 with a series of trial irradiations which were performed in close collaboration with Wacker Chemitronic of Germany. A methodology for the determination of irradiation times necessary to achieve the target resistivities was verified on the basis of the results from the trial irradiations. All production activitiesmore » are controlled by quality assurance procedures. To date some hundred and twelve silicon ingots (103 mm diameter) have been successfully irradiated on a commercial contract basis. The observed axial and radial variations in the resistivity profile of the ingots are very small compared to the profiles associated with conventionally doped silicon and small tolerances on target resistivities are attained. In this paper an overview of the design and characterisation of SILIRAD is given and the methods applied that ensure a quality product are described. Results obtained from trial and production irradiations are presented and the envisaged future modifications to SILIRAD discussed.« less

  7. Present status of liquid metal research for a fusion reactor

    NASA Astrophysics Data System (ADS)

    Tabarés, Francisco L.

    2016-01-01

    Although the use of solid materials as targets of divertor plasmas in magnetic fusion research is accepted as the standard solution for the very challenging issue of power and particle handling in a fusion reactor, a generalized feeling that the present options chosen for ITER will not represent the best choice for a reactor is growing up. The problems found for tungsten, the present selection for the divertor target of ITER, in laboratory tests and in hot plasma fusion devices suggest so. Even in the absence of the strong neutron irradiation expected in a reactor, issues like surface melting, droplet ejection, surface cracking, dust generation, etc., call for alternative solutions in a long pulse, high efficient fusion energy-producing continuous machine. Fortunately enough, decades of research on plasma facing materials based on liquid metals (LMs) have produced a wealth of appealing ideas that could find practical application in the route to the realization of a commercial fusion power plant. The options presently available, although in a different degree of maturity, range from full coverage of the inner wall of the device with liquid metals, so that power and particle exhaust together with neutron shielding could be provided, to more conservative combinations of liquid metal films and conventional solid targets basically representing a sort of high performance, evaporative coating for the alleviation of the surface degradation issues found so far. In this work, an updated review of worldwide activities on LM research is presented, together with some open issues still remaining and some proposals based on simple physical considerations leading to the optimization of the most conservative alternatives.

  8. NEUTRONIC REACTOR

    DOEpatents

    Hurwitz, H. Jr.; Brooks, H.; Mannal, C.; Payne, J.H.; Luebke, E.A.

    1959-03-24

    A reactor of the heterogeneous, liquid cooled type is described. This reactor is comprised of a central region of a plurality of vertically disposed elongated tubes surrounded by a region of moderator material. The central region is comprised of a central core surrounded by a reflector region which is surrounded by a fast neutron absorber region, which in turn is surrounded by a slow neutron absorber region. Liquid sodium is used as the primary coolant and circulates through the core which contains the fuel elements. Control of the reactor is accomplished by varying the ability of the reflector region to reflect neutrons back into the core of the reactor. For this purpose the reflector is comprised of moderator and control elements having varying effects on reactivity, the control elements being arranged and actuated by groups to give regulation, shim, and safety control.

  9. Overview of the Westinghouse Small Modular Reactor building layout

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Cronje, J. M.; Van Wyk, J. J.; Memmott, M. J.

    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 themore » 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

  10. Nuclear Safety. Technical progress journal, January--March 1994: Volume 35, No. 1

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Silver, E G

    1994-01-01

    This is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities.

  11. Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Petrovic, Bojan; Maldonado, Ivan

    2016-04-14

    The research performed in this project addressed the issue of low heavy metal loading and the resulting reduced cycle length with increased refueling frequency, inherent to all FHR designs with solid, non-movable fuel based on TRISO particles. Studies performed here focused on AHTR type of reactor design with plate (“plank”) fuel. Proposal to FY12 NEUP entitled “Fuel and Core Design Options to Overcome the Heavy Metal Loading Limit and Improve Performance and Safety of Liquid Salt Cooled Reactors” was selected for award, and the 3-year project started in August 2012. A 4-month NCE was granted and the project completed onmore » December 31, 2015. The project was performed by Georgia Tech (Prof. Bojan Petrovic, PI) and University of Tennessee (Prof. Ivan Maldonado, Co-PI), with a total funding of $758,000 over 3 years. In addition to two Co-PIs, the project directly engaged 6 graduate students (at doctoral or MS level) and 2 postdoctoral researchers. Additionally, through senior design projects and graduate advanced design projects, another 23 undergraduate and 12 graduate students were exposed to and trained in the salt reactor technology. We see this as one of the important indicators of the project’s success and effectiveness. In the process, 1 journal article was published (with 3 journal articles in preparation), together with 8 peer-reviewed full conference papers, 8 peer-reviewed extended abstracts, as well as 1 doctoral dissertation and 2 master theses. The work included both development of models and methodologies needed to adequately analyze this type of reactor, fuel, and its fuel cycle, as well as extensive analyses and optimization of the fuel and core design.« less

  12. Impact of Active Control on Passive Safety Response Characteristics of Sodium-Cooled Fast Reactors: II-Model Implementation and Simulations

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Ponciroli, Roberto; Passerini, Stefano; Vilim, Richard B.

    Advanced reactors are often claimed to be passively safe against unprotected upset events. In common practice, these events are not considered in the context of the plant control system, i.e., the reactor is subjected to classes of unprotected upset events while the normally programmed response of the control system is assumed not to be present. However, this approach constitutes an oversimplification since, depending on the upset involving the control system, an actuator does not necessarily go in the same direction as needed for safety. In this work, dynamic simulations are performed to assess the degree to which the inherent self-regulatingmore » plant response is safe from active control system override. The simulations are meant to characterize the resilience of the plant to unprotected initiators. The initiators were represented and modeled as an actuator going to a hard limit. Consideration of failure is further limited to individual controllers as there is no cross-connect of signals between these controllers. The potential for passive safety override by the control system is then relegated to the single-input single-output controllers. Here, the results show that when the plant control system is designed by taking into account and quantifying the impact of the plant control system on accidental scenarios there is very limited opportunity for the preprogrammed response of the control system to override passive safety protection in the event of an unprotected initiator.« less

  13. Impact of Active Control on Passive Safety Response Characteristics of Sodium-Cooled Fast Reactors: II-Model Implementation and Simulations

    DOE PAGES

    Ponciroli, Roberto; Passerini, Stefano; Vilim, Richard B.

    2017-06-21

    Advanced reactors are often claimed to be passively safe against unprotected upset events. In common practice, these events are not considered in the context of the plant control system, i.e., the reactor is subjected to classes of unprotected upset events while the normally programmed response of the control system is assumed not to be present. However, this approach constitutes an oversimplification since, depending on the upset involving the control system, an actuator does not necessarily go in the same direction as needed for safety. In this work, dynamic simulations are performed to assess the degree to which the inherent self-regulatingmore » plant response is safe from active control system override. The simulations are meant to characterize the resilience of the plant to unprotected initiators. The initiators were represented and modeled as an actuator going to a hard limit. Consideration of failure is further limited to individual controllers as there is no cross-connect of signals between these controllers. The potential for passive safety override by the control system is then relegated to the single-input single-output controllers. Here, the results show that when the plant control system is designed by taking into account and quantifying the impact of the plant control system on accidental scenarios there is very limited opportunity for the preprogrammed response of the control system to override passive safety protection in the event of an unprotected initiator.« less

  14. DYNAMIC AND STATIC PARAMETERS OF THE AQUEOUS HOMOGENEOUS ARMOUR RESEARCH REACTOR

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Terrell, C.W.; McElroy, W.N.

    1959-06-01

    A brief description of the aqueous homogeneous Armour Research Reactor is given. The negative reactivity coefficient resulting from a temperature increase was determined over a fuel temperature range of 37 to 150 deg F. Possession of an accurately calibrated rod and temperature coefficient permitted a direct measurement of the void coefficient. The reactor was taken to different power levels, and from the calibrated rod the total reduction in excess reactivity was obtained. During the power increase program additional U/sup 235/ and water were added to the core to determine the worth of U/sup 235/ and water. (W.D.M.)

  15. The combined hybrid system: A symbiotic thermal reactor/fast reactor system for power generation and radioactive waste toxicity reduction

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Hollaway, W.R.

    1991-08-01

    If there is to be a next generation of nuclear power in the United States, then the four fundamental obstacles confronting nuclear power technology must be overcome: safety, cost, waste management, and proliferation resistance. The Combined Hybrid System (CHS) is proposed as a possible solution to the problems preventing a vigorous resurgence of nuclear power. The CHS combines Thermal Reactors (for operability, safety, and cost) and Integral Fast Reactors (for waste treatment and actinide burning) in a symbiotic large scale system. The CHS addresses the safety and cost issues through the use of advanced reactor designs, the waste management issuemore » through the use of actinide burning, and the proliferation resistance issue through the use of an integral fuel cycle with co-located components. There are nine major components in the Combined Hybrid System linked by nineteen nuclear material mass flow streams. A computer code, CHASM, is used to analyze the mass flow rates CHS, and the reactor support ratio (the ratio of thermal/fast reactors), IFR of the system. The primary advantages of the CHS are its essentially actinide-free high-level radioactive waste, plus improved reactor safety, uranium utilization, and widening of the option base. The primary disadvantages of the CHS are the large capacity of IFRs required (approximately one MW{sub e} IFR capacity for every three MW{sub e} Thermal Reactor) and the novel radioactive waste streams produced by the CHS. The capability of the IFR to burn pure transuranic fuel, a primary assumption of this study, has yet to be proven. The Combined Hybrid System represents an attractive option for future nuclear power development; that disposal of the essentially actinide-free radioactive waste produced by the CHS provides an excellent alternative to the disposal of intact actinide-bearing Light Water Reactor spent fuel (reducing the toxicity based lifetime of the waste from roughly 360,000 years to about 510 years).« less

  16. Reactor shutdown delays medical procedures

    NASA Astrophysics Data System (ADS)

    Gwynne, Peter

    2008-01-01

    A longer-than-expected maintenance shutdown of the Canadian nuclear reactor that produces North America's entire supply of molybdenum-99 - from which the radioactive isotopes technetium-99 and iodine-131 are made - caused delays to the diagnosis and treatment of thousands of seriously ill patients last month. Technetium-99 is a key component of nuclear-medicine scans, while iodine-131 is used to treat cancer and other diseases of the thyroid. Production eventually resumed, but only after the Canadian government had overruled the Canadian Nuclear Safety Commission (CNSC), which was still concerned about the reactor's safety.

  17. Advanced research workshop: nuclear materials safety

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Jardine, L J; Moshkov, M M

    The Advanced Research Workshop (ARW) on Nuclear Materials Safety held June 8-10, 1998, in St. Petersburg, Russia, was attended by 27 Russian experts from 14 different Russian organizations, seven European experts from six different organizations, and 14 U.S. experts from seven different organizations. The ARW was conducted at the State Education Center (SEC), a former Minatom nuclear training center in St. Petersburg. Thirty-three technical presentations were made using simultaneous translations. These presentations are reprinted in this volume as a formal ARW Proceedings in the NATO Science Series. The representative technical papers contained here cover nuclear material safety topics on themore » storage and disposition of excess plutonium and high enriched uranium (HEU) fissile materials, including vitrification, mixed oxide (MOX) fuel fabrication, plutonium ceramics, reprocessing, geologic disposal, transportation, and Russian regulatory processes. This ARW completed discussions by experts of the nuclear materials safety topics that were not covered in the previous, companion ARW on Nuclear Materials Safety held in Amarillo, Texas, in March 1997. These two workshops, when viewed together as a set, have addressed most nuclear material aspects of the storage and disposition operations required for excess HEU and plutonium. As a result, specific experts in nuclear materials safety have been identified, know each other from their participation in t he two ARW interactions, and have developed a partial consensus and dialogue on the most urgent nuclear materials safety topics to be addressed in a formal bilateral program on t he subject. A strong basis now exists for maintaining and developing a continuing dialogue between Russian, European, and U.S. experts in nuclear materials safety that will improve the safety of future nuclear materials operations in all the countries involved because of t he positive synergistic effects of focusing these diverse backgrounds

  18. A neutronics feasibility study for the LEU conversion of Poland's Maria research reactor.

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Bretscher, M. M.

    1998-10-14

    The MARIA reactor is a high-flux multipurpose research reactor which is water-cooled and moderated with both beryllium and water. Standard HEU (80% {sup 235}U)fuel assemblies consist of six concentric fuel tubes of a U-Al alloy clad in aluminum. Although the inventory of HEU (80%) fuel is nearly exhausted, a supply of highly-loaded 36%-enriched fuel assemblies is available at the reactor site. Neutronic equilibrium studies have been made to determine the relative performance of fuels with enrichments of 80%, 36% and 19.7%. These studies indicate that LEU (19.7%) densities of about 2.5 gU/cm{sup 3} and 3.8 gU/cm{sup 3} are required tomore » match the performance of the MARIA reactor with 80%-enriched and with 36%-enriched fuels, respectively.« less

  19. Double-clad nuclear fuel safety rod

    DOEpatents

    McCarthy, William H.; Atcheson, Donald B.; Vaidyanathan, Swaminathan

    1984-01-01

    A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

  20. Double-clad nuclear-fuel safety rod

    DOEpatents

    McCarthy, W.H.; Atcheson, D.B.

    1981-12-30

    A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

  1. Research and technology in the Federal Motor Carrier Safety Administration

    DOT National Transportation Integrated Search

    2002-01-01

    As the Federal Government's chief commercial vehicle safety agency, the Federal Motor Carrier Safety Administration's (FMCSA), Office of Research and Technology (R&T) focuses on saving lives and reducing injuries by helping to prevent crashes involvi...

  2. Risk Management for Sodium Fast Reactors.

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Denman, Matthew R.; Groth, Katrina; Cardoni, Jeffrey N.

    2015-01-01

    Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self - correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the syste m's design to manage the accident. While inherently and passively safe designs are laudable, extreme boundary conditions can interfere with the design attributes which facilitate inherent safety , thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event withmore » the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayes ian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The author s would like to acknowledge the U.S. Department of E nergy's Office of Nuclear Energy for funding this research through Work Package SR - 14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at A rgonne N ational L aborator y , O ak R idge N ational L aborator y , and I daho N ational L aborator y for their continue d contributions to the advanced reactor PRA mission area.« less

  3. Identifying research priorities for patient safety in mental health: an international expert Delphi study

    PubMed Central

    Murray, Kevin; Thibaut, Bethan; Ramtale, Sonny Christian; Adam, Sheila; Darzi, Ara; Archer, Stephanie

    2018-01-01

    Objective Physical healthcare has dominated the patient safety field; research in mental healthcare is not as extensive but findings from physical healthcare cannot be applied to mental healthcare because it delivers specialised care that faces unique challenges. Therefore, a clearer focus and recognition of patient safety in mental health as a distinct research area is still needed. The study aim is to identify future research priorities in the field of patient safety in mental health. Design Semistructured interviews were conducted with the experts to ascertain their views on research priorities in patient safety in mental health. A three-round online Delphi study was used to ascertain consensus on 117 research priority statements. Setting and participants Academic and service user experts from the USA, UK, Switzerland, Netherlands, Ireland, Denmark, Finland, Germany, Sweden, Australia, New Zealand and Singapore were included. Main outcome measures Agreement in research priorities on a five-point scale. Results Seventy-nine statements achieved consensus (>70%). Three out of the top six research priorities were patient driven; experts agreed that understanding the patient perspective on safety planning, on self-harm and on medication was important. Conclusions This is the first international Delphi study to identify research priorities in safety in the mental field as determined by expert academic and service user perspectives. A reasonable consensus was obtained from international perspectives on future research priorities in patient safety in mental health; however, the patient perspective on their mental healthcare is a priority. The research agenda for patient safety in mental health identified here should be informed by patient safety science more broadly and used to further establish this area as a priority in its own right. The safety of mental health patients must have parity with that of physical health patients to achieve this. PMID:29502096

  4. Analysis of UF6 breeder reactor power plants

    NASA Technical Reports Server (NTRS)

    Clement, J. D.; Rust, J. H.

    1976-01-01

    Gaseous UF6 fueled breeder reactor design and technical applications of such concepts are summarized. Special attention was given to application in nuclear power plants and to reactor efficiency and safety factors.

  5. Safety design approach for external events in Japan sodium-cooled fast reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Yamano, H.; Kubo, S.; Tani, A.

    2012-07-01

    This paper describes a safety design approach for external events in the design study of Japan sodium-cooled fast reactor. An emphasis is introduction of a design extension external condition (DEEC). In addition to seismic design, other external events such as tsunami, strong wind, abnormal temperature, etc. were addressed in this study. From a wide variety of external events consisting of natural hazards and human-induced ones, a screening method was developed in terms of siting, consequence, frequency to select representative events. Design approaches for these events were categorized on the probabilistic, statistical and deterministic basis. External hazard conditions were considered mainlymore » for DEECs. In the probabilistic approach, the DEECs of earthquake, tsunami and strong wind were defined as 1/10 of exceedance probability of the external design bases. The other representative DEECs were also defined based on statistical or deterministic approaches. (authors)« less

  6. RADON LEVELS AND ЕQUIVALENT DOSE RATES AT THE IRT-SOFIA RESEARCH REACTOR SITE.

    PubMed

    Krezhov, Kiril; Mladenov, Aleksander; Dimitrov, Dobromir

    2018-06-11

    Results from radon measurements by active sampling of indoor air in the buildings within the Nuclear Scientific Experimental and Educational Centre (NSEEC) protected site at the Institute for Nuclear Research and Nuclear Energy (INRNE) are presented. The inspected buildings included in this report are the IRT research reactor structure and several auxiliary formations wherein the laundry facilities and the gamma irradiator GOU-1 (60Co source) are installed as well as the Central Alarm Station (CAS) premises. Besides the reactor hall and the primary cooling loop area, special attention was given to the premises of the First Class Radiochemical Laboratory in the IRT reactor basement. Determination of radon concentration distribution in the premises of the constructions within the site is an important part of radiation surveillance during the operation and maintenance of the NSEEC facilities as well as for their involvement in the educational activities at INRNE.

  7. Dismantlement of the TSF-SNAP Reactor Assembly

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Peretz, Fred J

    2009-01-01

    This paper describes the dismantlement of the Tower Shielding Facility (TSF)?Systems for Nuclear Auxiliary Power (SNAP) reactor, a SNAP-10A reactor used to validate radiation source terms and shield performance models at Oak Ridge National Laboratory (ORNL) from 1967 through 1973. After shutdown, it was placed in storage at the Y-12 National Security Complex (Y-12), eventually falling under the auspices of the Highly Enriched Uranium (HEU) Disposition Program. To facilitate downblending of the HEU present in the fuel elements, the TSF-SNAP was moved to ORNL on June 24, 2006. The reactor assembly was removed from its packaging, inspected, and the sodium-potassiummore » (NaK) coolant was drained. A superheated steam process was used to chemically react the residual NaK inside the reactor assembly. The heat exchanger assembly was removed from the top of the reactor vessel, and the criticality safety sleeve was exchanged for a new safety sleeve that allowed for the removal of the vessel lid. A chain-mounted tubing cutter was used to separate the lid from the vessel, and the 36 fuel elements were removed and packaged in four U.S. Department of Transportation 2R/6M containers. The fuel elements were returned to Y-12 on July 13, 2006. The return of the fuel elements and disposal of all other reactor materials accomplished the formal objectives of the dismantlement project. In addition, a project model was established for the handling of a fully fueled liquid-metal?cooled reactor assembly. Current criticality safety codes have been benchmarked against experiments performed by Atomics International in the 1950s and 1960s. Execution of this project provides valuable experience applicable to future projects addressing space and liquid-metal-cooled reactors.« less

  8. Georgia Institute of Technology research on the Gas Core Actinide Transmutation Reactor (GCATR)

    NASA Technical Reports Server (NTRS)

    Clement, J. D.; Rust, J. H.; Schneider, A.; Hohl, F.

    1976-01-01

    The program reviewed is a study of the feasibility, design, and optimization of the GCATR. The program is designed to take advantage of initial results and to continue work carried out on the Gas Core Breeder Reactor. The program complements NASA's program of developing UF6 fueled cavity reactors for power, nuclear pumped lasers, and other advanced technology applications. The program comprises: (1) General Studies--Parametric survey calculations performed to examine the effects of reactor spectrum and flux level on the actinide transmutation for GCATR conditions. The sensitivity of the results to neutron cross sections are to be assessed. Specifically, the parametric calculations of the actinide transmutation are to include the mass, isotope composition, fission and capture rates, reactivity effects, and neutron activity of recycled actinides. (2) GCATR Design Studies--This task is a major thrust of the proposed research program. Several subtasks are considered: optimization criteria studies of the blanket and fuel reprocessing, the actinide insertion and recirculation system, and the system integration. A brief review of the background of the GCATR and ongoing research is presented.

  9. 76 FR 17808 - Final Vehicle Safety Rulemaking and Research Priority Plan 2011-2013

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-31

    ... [Docket No. NHTSA-2009-0108] Final Vehicle Safety Rulemaking and Research Priority Plan 2011- 2013 AGENCY... availability. SUMMARY: This document announces the availability of the Final NHTSA Vehicle Safety and Fuel.... This Priority Plan is an update to the Final Vehicle Safety Rulemaking and Research Priority Plan 2009...

  10. RAZORBACK - A Research Reactor Transient Analysis Code Version 1.0 - Volume 3: Verification and Validation Report.

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Talley, Darren G.

    2017-04-01

    This report describes the work and results of the verification and validation (V&V) of the version 1.0 release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, the equation of motion for fuel element thermal expansion, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This V&V effort was intended to confirm that the code showsmore » good agreement between simulation and actual ACRR operations.« less

  11. The National Shipbuilding Research Program: Employee Involvement/Safety

    DTIC Science & Technology

    1990-06-01

    THE NATIONAL SHIPBUILDING RESEARCH PROGRAM Employee InvoIvement/Safety U.S. DEPARTMENT OF TRANSPORTATION Maritime Administration and U.S. NAVY in...to and sought assistance either directly or through the Program Manager or the MTC Safety Chair- man from individual members who had functional respon...carpenters in the Model Shop. The training the2. 3. 4. 5. program to be developed and taught by the SP-5 Team. (The employees in the Model Shop were selected

  12. Heat Transfer Salts for Nuclear Reactor Systems - Chemistry Control, Corrosion Mitigation, and Modeling

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Anderson, Mark; Sridharan, Kumar; Morgan, Dane

    2015-01-22

    The concept of a molten salt reactor has existed for nearly sixty years. Previously all work was done during a large collaborative effort at Oak Ridge National Laboratory, culminating in a research reactor which operated for 15,000 hours without major error. This technical success has garnished interest in modern, high temperature, reactor schemes. Research using molten fluoride salts for nuclear applications requires a steady supply of high grade molten salts. There is no bulk supplier of research grade fluoride salts in the world, so a facility which could provide all the salt needed for testing at the University of Wisconsinmore » had to be produced. Two salt purification devices were made for this purpose, a large scale purifier, and a small scale purifier, each designed to clean the salts from impurities and reduce their corrosion potential. As of now, the small scale has performed with flibe salt, hydrogen, and hydrogen fluoride, yielding clean salt. This salt is currently being used in corrosion testing facilities at the Massachusetts Institute of Technology and the University of Wisconsin. Working with the beryllium based salts requires extensive safety measures and health monitoring to prevent the development of acute or chronic beryllium disease, two pulmonary diseases created by an allergic reaction to beryllium in the lungs. Extensive health monitoring, engineering controls, and environment monitoring had to be set up with the University of Wisconsin department of Environment, Health and Safety. The hydrogen fluoride required for purification was also an extreme health hazard requiring thoughtful planning and execution. These dangers have made research a slow and tedious process. Simple processes, such as chemical handling and clean-up, can take large amounts of ingenuity and time. Other work has complemented the experimental research at Wisconsin to advance high temperature reactor goals. Modeling work has been performed in house to re

  13. Conducting Clinically Based Intimate Partner Violence Research: Safety Protocol Recommendations.

    PubMed

    Anderson, Jocelyn C; Glass, Nancy E; Campbell, Jacquelyn C

    Maintaining safety is of utmost importance during research involving participants who have experienced intimate partner violence (IPV). Limited guidance on safety protocols to protect participants is available, particularly information related to technology-based approaches to informed consent, data collection, and contacting participants during the course of a study. The purpose of the article is to provide details on the safety protocol developed and utilized with women receiving care at an urban HIV clinic and who were taking part in an observational study of IPV, mental health symptoms, and substance abuse and their relationship to HIV treatment adherence. The protocol presents the technological strategies to promote safety and allow autonomy in participant decision-making throughout the research process, including Voice over Internet Protocol telephone numbers, and tablet-based eligibility screening and data collection. Protocols for management of participants at risk for suicide and/or intimate partner homicide that included automated high-risk messaging to participants and research staff and facilitated disclosure of risk to clinical staff based on participant preferences are discussed. Use of technology and partnership with clinic staff helped to provide an environment where research regarding IPV could be conducted without undue burden or risk to participants. Utilizing tablet-based survey administration provided multiple practical and safety benefits for participants. Most women who screened into high-risk categories for suicide or intimate partner homicide did not choose to have their results shared with their healthcare providers, indicating the importance of allowing participants control over information sharing whenever possible.

  14. Nuclear power: Siting and safety

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Openshaw, S.

    1986-01-01

    By 2030, half, or even two-thirds, of all electricity may be generated by nuclear power. Major reactor accidents are still expected to be rare occurrences, but nuclear safety is largely a matter of faith. Terrorist attacks, sabotage, and human error could cause a significant accident. Reactor siting can offer an additional, design-independent margin of safety. Remote geographical sites for new plants would minimize health risks, protect the industry from negative changes in public opinion concerning nuclear energy, and improve long-term public acceptance of nuclear power. U.K. siting practices usually do not consider the contribution to safety that could be obtainedmore » from remote sites. This book discusses the present trends of siting policies of nuclear power and their design-independent margin of safety.« less

  15. An Innovative Hybrid Loop-Pool SFR Design and Safety Analysis Methods: Today and Tomorrow

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Hongbin Zhang; Haihua Zhao; Vincent Mousseau

    2008-04-01

    Investment in commercial sodium cooled fast reactor (SFR) power plants will become possible only if SFRs achieve economic competitiveness as compared to light water reactors and other Generation IV reactors. Toward that end, we have launched efforts to improve the economics and safety of SFRs from the thermal design and safety analyses perspectives at Idaho National Laboratory. From the thermal design perspective, an innovative hybrid loop-pool SFR design has been proposed. This design takes advantage of the inherent safety of a pool design and the compactness of a loop design to further improve economics and safety. From the safety analysesmore » perspective, we have initiated an effort to develop a high fidelity reactor system safety code.« less

  16. Methods and codes for neutronic calculations of the MARIA research reactor.

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Andrzejewski, K.; Kulikowska, T.; Bretscher, M. M.

    2002-02-18

    The core of the MARIA high flux multipurpose research reactor is highly heterogeneous. It consists of beryllium blocks arranged in 6 x 8 matrix, tubular fuel assemblies, control rods and irradiation channels. The reflector is also heterogeneous and consists of graphite blocks clad with aluminum. Its structure is perturbed by the experimental beam tubes. This paper presents methods and codes used to calculate the MARIA reactor neutronics characteristics and experience gained thus far at IAE and ANL. At ANL the methods of MARIA calculations were developed in connection with the RERTR program. At IAE the package of programs was developedmore » to help its operator in optimization of fuel utilization.« less

  17. Identifying research priorities for patient safety in mental health: an international expert Delphi study.

    PubMed

    Dewa, Lindsay H; Murray, Kevin; Thibaut, Bethan; Ramtale, Sonny Christian; Adam, Sheila; Darzi, Ara; Archer, Stephanie

    2018-03-03

    Physical healthcare has dominated the patient safety field; research in mental healthcare is not as extensive but findings from physical healthcare cannot be applied to mental healthcare because it delivers specialised care that faces unique challenges. Therefore, a clearer focus and recognition of patient safety in mental health as a distinct research area is still needed. The study aim is to identify future research priorities in the field of patient safety in mental health. Semistructured interviews were conducted with the experts to ascertain their views on research priorities in patient safety in mental health. A three-round online Delphi study was used to ascertain consensus on 117 research priority statements. Academic and service user experts from the USA, UK, Switzerland, Netherlands, Ireland, Denmark, Finland, Germany, Sweden, Australia, New Zealand and Singapore were included. Agreement in research priorities on a five-point scale. Seventy-nine statements achieved consensus (>70%). Three out of the top six research priorities were patient driven; experts agreed that understanding the patient perspective on safety planning, on self-harm and on medication was important. This is the first international Delphi study to identify research priorities in safety in the mental field as determined by expert academic and service user perspectives. A reasonable consensus was obtained from international perspectives on future research priorities in patient safety in mental health; however, the patient perspective on their mental healthcare is a priority. The research agenda for patient safety in mental health identified here should be informed by patient safety science more broadly and used to further establish this area as a priority in its own right. The safety of mental health patients must have parity with that of physical health patients to achieve this. © Article author(s) (or their employer(s) unless otherwise stated in the text of the article) 2018. All

  18. Simulator platform for fast reactor operation and safety technology demonstration

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Vilim, R. B.; Park, Y. S.; Grandy, C.

    2012-07-30

    A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe responsemore » to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.« less

  19. Utilization of the Philippine Research Reactor as a training facility for nuclear power plant operators

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Palabrica, R.J.

    1981-01-01

    The Philippines has a 1-MW swimming-pool reactor facility operated by the Philippine Atomic Energy Commission (PAEC). The reactor is light-water moderated and cooled, graphite reflected, and fueled with 90% enriched uranium. Since it became critical in 1963 it has been utilized for research, radioisotope production, and training. It was used initially in the training of PAEC personnel and other research institutions and universities. During the last few years, however, it has played a key role in training personnel for the Philippine Nuclear Power Project (PNPP).

  20. Overview of bureau research directed towards surface powered haulage safety

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    May, J.P.; Aldinger, J.A.

    1995-12-31

    Surface mining operations, including mills and preparation plants, employ over 260,000 people. This represents a significant contribution to our nation`s economy and an important source of skilled and well-paying jobs. As mine production has shifted from underground to surface, and with continuing advances in underground mine safety, surface mining has unfortunately become the leader in mine fatalities. In 1994 surface mining accidents accounted for 49% of all mine fatalities, followed by underground mining with 37% and mills and preparation plants with 14%. The U.S. Bureau of Mines (USBM) has targeted surface mining as an important research priority to reduce themore » social and economic costs associated with fatalities and lost-work-time injuries. USBM safety research focuses on the development of technologies that can enhance productivity and reduce mining costs through a reduction in the number and severity of mining accidents. This report summarizes a number of completed and ongoing research programs directed towards surface powered haulage--the single largest category of fatalities in surface mining and a major cause of lost workdays. Research products designed for industry are highlighted and future USBM surface mining safety research is discussed.« less

  1. Multi-physics design and analyses of long life reactors for lunar outposts

    NASA Astrophysics Data System (ADS)

    Schriener, Timothy M.

    Future human exploration of the solar system is likely to include establishing permanent outposts on the surface of the Moon. These outposts will require reliable sources of electrical power in the range of 10's to 100's of kWe to support exploration and resource utilization activities. This need is best met using nuclear reactor power systems which can operate steadily throughout the long ˜27.3 day lunar rotational period, irrespective of location. Nuclear power systems can potentially open up the entire lunar surface for future exploration and development. Desirable features of nuclear power systems for the lunar surface include passive operation, the avoidance of single point failures in reactor cooling and the integrated power system, moderate operating temperatures to enable the use of conventional materials with proven irradiation experience, utilization of the lunar regolith for radiation shielding and as a supplemental neutron reflector, and safe post-operation decay heat removal and storage for potential retrieval. In addition, it is desirable for the reactor to have a long operational life. Only a limited number of space nuclear reactor concepts have previously been developed for the lunar environment, and these designs possess only a few of these desirable design and operation features. The objective of this research is therefore to perform design and analyses of long operational life lunar reactors and power systems which incorporate the desirable features listed above. A long reactor operational life could be achieved either by increasing the amount of highly enriched uranium (HEU) fuel in the core or by improving the neutron economy in the reactor through reducing neutron leakage and parasitic absorption. The amount of fuel in surface power reactors is constrained by the launch safety requirements. These include ensuring that the bare reactor core remains safely subcritical when submerged in water or wet sand and flooded with seawater in the unlikely

  2. Proceedings of the 1988 International Meeting on Reduced Enrichment for Research and Test Reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Not Available

    1993-07-01

    The international effort to develop and implement new research reactor fuels utilizing low-enriched uranium, instead of highly- enriched uranium, continues to make solid progress. This effort is the cornerstone of a widely shared policy aimed at reducing, and possibly eliminating, international traffic in highly-enriched uranium and the nuclear weapon proliferation concerns associated with this traffic. To foster direct communication and exchange of ideas among the specialists in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the eleventh of a series which began 1978. Individual papers presented at the meetingmore » have been cataloged separately.« less

  3. D-He-3 spherical torus fusion reactor system study

    NASA Astrophysics Data System (ADS)

    Macon, William A., Jr.

    1992-04-01

    This system study extrapolates present physics knowledge and technology to predict the anticipated characteristics of D-He3 spherical torus fusion reactors and their sensitivity to uncertainties in important parameters. Reference cases for steady-state 1000 MWe reactors operating in H-mode in both the 1st stability regime and the 2nd stability regime were developed and assessed quantitatively. These devices would a very small aspect ratio (A=1,2), a major radius of about 2.0 m, an on-axis magnetic field less than 2 T, a large plasma current (80-120 MA) dominated by the bootstrap effect, and high plasma beta (greater than O.6). The estimated cost of electricity is in the range of 60-90 mills/kW-hr, assuming the use of a direct energy conversion system. The inherent safety and environmental advantages of D-He3 fusion indicate that this reactor concept could be competitive with advanced fission breeder reactors and large-scale solar electric plants by the end of the 21st century if research and development can produce the anticipated physics and technology advances.

  4. Nuclear Safety

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Silver, E G

    This document is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.

  5. Defining Safety in the Nursing Home Setting: Implications for Future Research.

    PubMed

    Simmons, Sandra F; Schnelle, John F; Sathe, Nila A; Slagle, Jason M; Stevenson, David G; Carlo, Maria E; McPheeters, Melissa L

    2016-06-01

    Currently, the Agency for Healthcare Research and Quality (AHRQ) Common Format for nursing homes (NHs) accommodates voluntary reporting for 4 adverse events: falls with injury, pressure ulcers, medication errors, and infections. In 2015, AHRQ funded a technical brief to describe the state of the science related to safety in the NH setting to inform a research agenda. Thirty-six recent systematic reviews evaluated NH safety-related interventions to address these 4 adverse events and reported mostly mixed evidence about effective approaches to ameliorate them. Furthermore, these 4 events are likely inadequate to capture safety issues that are unique to the NH setting and encompass other domains related to residents' quality of care and quality of life. Future research needs include expanding our definition of safety in the NH setting, which differs considerably from that of hospitals, to include contributing factors to adverse events as well as more resident-centered care measures. Second, future research should reflect more rigorous implementation science to include objective measures of care processes related to adverse events, intervention fidelity, and staffing resources for intervention implementation to inform broader uptake of efficacious interventions. Weaknesses in implementation contribute to the current inconclusive and mixed evidence base as well as remaining questions about what outcomes are even achievable in the NH setting, given the complexity of most resident populations. Also related to implementation, future research should determine the effects of specific staffing models on care processes related to safety outcomes. Last, future efforts should explore the potential for safety issues in other care settings for older adults, most notably dementia care within assisted living. Copyright © 2016 AMDA – The Society for Post-Acute and Long-Term Care Medicine. All rights reserved.

  6. Dissemination and Implementation Research for Occupational Safety and Health.

    PubMed

    Dugan, Alicia G; Punnett, Laura

    2017-12-01

    The translation of evidence-based health innovations into real-world practice is both incomplete and exceedingly slow. This represents a poor return on research investment dollars for the general public. U.S. funders of health sciences research (e.g., NIH, CDC, NIOSH) are increasingly calling for dissemination plans, and to a lesser extent for dissemination and implementation (D&I) research, which are studies that examine the effectiveness of D&I efforts and strategies and the predictors of D&I success. For example, rather than merely broadcasting information about a preventable hazard, D&I research in occupational safety and health (OSH) might examine how employers or practitioners are most likely to receive and act upon that information. We propose here that D&I research should be seen as a dedicated and necessary area of study within OSH, as a way to generate new knowledge that can bridge the research-to-practice gap. We present D&I concepts, frameworks, and examples that can increase the capacity of OSH professionals to conduct D&I research and accelerate the translation of research findings into meaningful everyday practice to improve worker safety and health.

  7. Research, development, and implementation of pedestrian safety facilities in the United Kingdom

    DOT National Transportation Integrated Search

    1999-12-01

    This report is one in a series of pedestrian safety synthesis reports prepared for the Federal Highway Administration to document pedestrian safety in other countries. This report reviews recent research on pedestrian safety carried out in the United...

  8. Reprocessing of research reactor fuel the Dounreay option

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Cartwright, P.

    1997-08-01

    Reprocessing is a proven process for the treatment of spent U/Al Research Reactor fuel. At Dounreay 12679 elements have been reprocessed during the past 30 years. For reactors converting to LEU fuel the uranium recovered in reprocessing can be blended down to less than 20% U{sub 235}, enrichment and be fabricated into new elements. For reactors already converted to LEU it is technically possible to reprocess spent silicide fuel to reduce the U{sub 235} burden and present to a repository only stable conditioned waste. The main waste stream from reprocessing which contains the Fission products is collected in underground storagemore » tanks where it is kept for a period of at least five years before being converted to a stable solid form for return to the country of origin for subsequent storage/disposal. Discharges to the environment from reprocessing are low and are limited to the radioactive gases contained in the spent fuel and a low level liquid waste steam. Both of these discharges are independently monitored, and controlled within strict discharge limits set by the UK Government`s Scottish Office. Transportation of spent fuel to Dounreay has been undertaken using many routes from mainland Europe and has utilised over the past few years both chartered and scheduled vessel services. Several different transport containers have been handled and are currently licensed in the UK. This paper provides a short history of MTR reprocessing at Dounreay, and provides information to show reprocessing can satisfy the needs of MTR operators, showing that reprocessing is a valuable asset in non-proliferation terms, offers a complete solution and is environmentally acceptable.« less

  9. NUCLEAR REACTOR

    DOEpatents

    Christy, R.F.

    1958-07-15

    A nuclear reactor of the homogeneous liquid fuel type is described wherein the fissionable isotope is suspended or dissolved in a liquid moderator such as water. The reactor core is comprised essentially of a spherical vessel for containing the reactive composition surrounded by a reflector, preferably of beryllium oxide. The reactive composition may be an ordinary water solution of a soluble salt of uranium, the quantity of fissionable isotope in solution being sufficient to provide a critical mass in the vessel. The liquid fuel is stored in a tank of non-crtttcal geometry below the reactor vessel and outside of the reflector and is passed from the tank to the vessel through a pipe connecting the two by air pressure means. Neutron absorbing control and safety rods are operated within slots in the reflector adjacent to the vessel.

  10. METHOD AND APPARATUS FOR REACTOR SAFETY CONTROL

    DOEpatents

    Huston, N.E.

    1961-06-01

    A self-contained nuclear reactor fuse controlled device tron absorbing material, normally in a compact form but which can be expanded into an extended form presenting a large surface for neutron absorption when triggered by an increase in neutron flux, is described.

  11. A Review of Research on Driving Styles and Road Safety.

    PubMed

    Sagberg, Fridulv; Selpi; Piccinini, Giulio Francesco Bianchi; Engström, Johan

    2015-11-01

    The aim of this study was to outline a conceptual framework for understanding driving style and, on this basis, review the state-of-the-art research on driving styles in relation to road safety. Previous research has indicated a relationship between the driving styles adopted by drivers and their crash involvement. However, a comprehensive literature review of driving style research is lacking. A systematic literature search was conducted, including empirical, theoretical, and methodological research, on driving styles related to road safety. A conceptual framework was proposed whereby driving styles are viewed in terms of driving habits established as a result of individual dispositions as well as social norms and cultural values. Moreover, a general scheme for categorizing and operationalizing driving styles was suggested. On this basis, existing literature on driving styles and indicators was reviewed. Links between driving styles and road safety were identified and individual and sociocultural factors influencing driving style were reviewed. Existing studies have addressed a wide variety of driving styles, and there is an acute need for a unifying conceptual framework in order to synthesize these results and make useful generalizations. There is a considerable potential for increasing road safety by means of behavior modification. Naturalistic driving observations represent particularly promising approaches to future research on driving styles. Knowledge about driving styles can be applied in programs for modifying driver behavior and in the context of usage-based insurance. It may also be used as a means for driver identification and for the development of driver assistance systems. © 2015, Human Factors and Ergonomics Society.

  12. Recent Upgrades at the Safety and Tritium Applied Research Facility

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Cadwallader, Lee Charles; Merrill, Brad Johnson; Stewart, Dean Andrew

    This paper gives a brief overview of the Safety and Tritium Applied Research (STAR) facility operated by the Fusion Safety Program (FSP) at the Idaho National Laboratory (INL). FSP researchers use the STAR facility to carry out experiments in tritium permeation and retention in various fusion materials, including wall armor tile materials. FSP researchers also perform other experimentation as well to support safety assessment in fusion development. This lab, in its present two-building configuration, has been in operation for over ten years. The main experiments at STAR are briefly described. This paper discusses recent work to enhance personnel safety atmore » the facility. The STAR facility is a Department of Energy less than hazard category 3 facility; the personnel safety approach calls for ventilation and tritium monitoring for radiation protection. The tritium areas of STAR have about 4 to 12 air changes per hour, with air flow being once through and then routed to the facility vent stack. Additional radiation monitoring has been installed to read the laboratory room air where experiments with tritium are conducted. These ion chambers and bubblers are used to verify that no significant tritium concentrations are present in the experiment rooms. Standby electrical power has been added to the facility exhaust blower so that proper ventilation will now operate during commercial power outages as well as the real-time tritium air monitors.« less

  13. Decommissioning of the High Flux Beam Reactor at Brookhaven National Laboratory.

    PubMed

    Hu, Jih-Perng; Reciniello, Richard N; Holden, Norman E

    2012-08-01

    The High Flux Beam Reactor (HFBR) at the Brookhaven National Laboratory was a heavy-water cooled and moderated reactor that achieved criticality on 31 October 1965. It operated at a power level of 40 mega-watts. An equipment upgrade in 1982 allowed operations at 60 mega-watts. After a 1989 reactor shutdown to reanalyze safety impact of a hypothetical loss of coolant accident, the reactor was restarted in 1991 at 30 mega-watts. The HFBR was shut down in December 1996 for routine maintenance and refueling. At that time, a leak of tritiated water was identified by routine sampling of ground water from wells located adjacent to the reactor's spent fuel pool. The reactor remained shut down for almost 3 y for safety and environmental reviews. In November 1999, the United States Department of Energy decided to permanently shut down the HFBR. The decontamination and decommissioning of the HFBR complex, consisting of multiple structures and systems to operate and maintain the reactor, were complete in 2009 after removing and shipping off all the control rod blades. The emptied and cleaned HFBR dome, which still contains the irradiated reactor vessel is presently under 24/7 surveillance for safety. Details of the HFBR's cleanup performed during 1999-2009, to allow the BNL facilities to be re-accessed by the public, will be described in the paper.

  14. Development of a Research Reactor Protocol for Neutron Multiplication Measurements

    DOE PAGES

    Arthur, Jennifer Ann; Bahran, Rian Mustafa; Hutchinson, Jesson D.; ...

    2018-03-20

    A new series of subcritical measurements has been conducted at the zero-power Walthousen Reactor Critical Facility (RCF) at Rensselaer Polytechnic Institute (RPI) using a 3He neutron multiplicity detector. The Critical and Subcritical 0-Power Experiment at Rensselaer (CaSPER) campaign establishes a protocol for advanced subcritical neutron multiplication measurements involving research reactors for validation of neutron multiplication inference techniques, Monte Carlo codes, and associated nuclear data. There has been increased attention and expanded efforts related to subcritical measurements and analyses, and this work provides yet another data set at known reactivity states that can be used in the validation of state-of-the-art Montemore » Carlo computer simulation tools. The diverse (mass, spatial, spectral) subcritical measurement configurations have been analyzed to produce parameters of interest such as singles rates, doubles rates, and leakage multiplication. MCNP ®6.2 was used to simulate the experiment and the resulting simulated data has been compared to the measured results. Comparison of the simulated and measured observables (singles rates, doubles rates, and leakage multiplication) show good agreement. This work builds upon the previous years of collaborative subcritical experiments and outlines a protocol for future subcritical neutron multiplication inference and subcriticality monitoring measurements on pool-type reactor systems.« less

  15. Development of a Research Reactor Protocol for Neutron Multiplication Measurements

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Arthur, Jennifer Ann; Bahran, Rian Mustafa; Hutchinson, Jesson D.

    A new series of subcritical measurements has been conducted at the zero-power Walthousen Reactor Critical Facility (RCF) at Rensselaer Polytechnic Institute (RPI) using a 3He neutron multiplicity detector. The Critical and Subcritical 0-Power Experiment at Rensselaer (CaSPER) campaign establishes a protocol for advanced subcritical neutron multiplication measurements involving research reactors for validation of neutron multiplication inference techniques, Monte Carlo codes, and associated nuclear data. There has been increased attention and expanded efforts related to subcritical measurements and analyses, and this work provides yet another data set at known reactivity states that can be used in the validation of state-of-the-art Montemore » Carlo computer simulation tools. The diverse (mass, spatial, spectral) subcritical measurement configurations have been analyzed to produce parameters of interest such as singles rates, doubles rates, and leakage multiplication. MCNP ®6.2 was used to simulate the experiment and the resulting simulated data has been compared to the measured results. Comparison of the simulated and measured observables (singles rates, doubles rates, and leakage multiplication) show good agreement. This work builds upon the previous years of collaborative subcritical experiments and outlines a protocol for future subcritical neutron multiplication inference and subcriticality monitoring measurements on pool-type reactor systems.« less

  16. Nuclear Reactor Physics

    NASA Astrophysics Data System (ADS)

    Stacey, Weston M.

    2001-02-01

    An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

  17. Loss-of-flow-without-scram tests in Experimental Breeder Reactor-II and comparison with pretest predictions

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Chang, L.K.; Mohr, D.; Planchon, H.P.

    This article discusses a series of successful loss-of-flow-without-scram tests conducted in Experimental Breeder Reactor-II (EBR-II), a metal-fueled, sodium-cooled fast reactor. These May 1985 tests demonstrated the capability of the EBR to reduce reactor power passively during a loss of flow and to maintain reactor temperatures within bounds without any reliance on an active safety system. The tests were run from reduced power to ensure that temperatures could be maintained well below the fuel-clad eutectic temperature. Good agreement was found between selected test data and pretest predictions made with the EBR-II system analysis code NATDEMO and the hot channel analysis codemore » HOTCHAN. The article also discusses safety assessments of the tests as well as modifications required on the EBR-II reactor safety system for conducting required on the EBR-II reactor safety system for the conducting the tests.« less

  18. 9 CFR 309.14 - Brucellosis-reactor goats.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 9 Animals and Animal Products 2 2011-01-01 2011-01-01 false Brucellosis-reactor goats. 309.14 Section 309.14 Animals and Animal Products FOOD SAFETY AND INSPECTION SERVICE, DEPARTMENT OF AGRICULTURE... INSPECTION AND CERTIFICATION ANTE-MORTEM INSPECTION § 309.14 Brucellosis-reactor goats. Goats which have...

  19. 9 CFR 309.14 - Brucellosis-reactor goats.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 9 Animals and Animal Products 2 2010-01-01 2010-01-01 false Brucellosis-reactor goats. 309.14 Section 309.14 Animals and Animal Products FOOD SAFETY AND INSPECTION SERVICE, DEPARTMENT OF AGRICULTURE... INSPECTION AND CERTIFICATION ANTE-MORTEM INSPECTION § 309.14 Brucellosis-reactor goats. Goats which have...

  20. A plasma arc reactor for fullerene research

    NASA Astrophysics Data System (ADS)

    Anderson, T. T.; Dyer, P. L.; Dykes, J. W.; Klavins, P.; Anderson, P. E.; Liu, J. Z.; Shelton, R. N.

    1994-12-01

    A modified Krätschmer-Huffman reactor for the mass production of fullerenes is presented. Fullerene mass production is fundamental for the synthesis of higher and endohedral fullerenes. The reactor employs mechanisms for continuous graphite-rod feeding and in situ slag removal. Soot collects into a Soxhlet extraction thimble which serves as a fore-line vacuum pump filter, thereby easing fullerene separation from soot. Thermal gravimetric analysis (TGA) for yield determination is reported. This TGA method is faster and uses smaller samples than Soxhlet extraction methods which rely on aromatic solvents. Production of 10 g of soot per hour is readily achieved utilizing this reactor. Fullerene yields of 20% are attained routinely.

  1. Automated power control system for reactor TRIGA PUSPATI

    NASA Astrophysics Data System (ADS)

    Ghazali, Anith Khairunnisa; Minhat, Mohd Sabri; Hassan, Mohd Khair

    2017-01-01

    Reactor TRIGA PUSPATI (RTP) Mark II type undergoes safe operation for more than 30 years and the only research reactor exists in Malaysia. The main safety feature of Instrumentation and Control (I&C) system design is such that any failure in the electronic, or its associated components, does not lead to an uncontrolled rate of reactivity. The existed controller using feedback approach to control the reactor power. This paper introduces proposed controllers such as Model Reference Adaptive Control (MRAC) and Proportional Integral Derivatives (PID) controller for the RTP simulation. In RTP, the most important considered parameter is the reactor power and act as nervous system. To design a controller for complex plant like RTP is quite difficult due to high cost and safety factors cause by the failure of the controller. Furthermore, to overcome these problems, a simulator can be used to replace functions the hardware and test could then be simulated using this simulator. In order to find the best controller, several controllers were proposed and the result will be analysed for study the performances of the controller. The output result will be used to find out the best RTP power controller using MATLAB/Simulink and gives result as close as the real RTP performances. Currently, the structures of RTP was design using MATLAB/Simulink tool that consist of fission chamber, controller, control rod position, height-to-worth of control rods and a RTP model. The controller will control the control rod position to make sure that the reactivity still under the limitation parameter. The results given from each controller will be analysed and validated through experiment data collected from RTP.

  2. Microprocessor tester for the treat upgrade reactor trip system

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Lenkszus, F.R.; Bucher, R.G.

    1984-01-01

    The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. In addition, a programmable Automated Reactor Control System (ARCS) will permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety systemmore » is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations.« less

  3. Design of a 25-kWe Surface Reactor System Based on SNAP Reactor Technologies

    NASA Astrophysics Data System (ADS)

    Dixon, David D.; Hiatt, Matthew T.; Poston, David I.; Kapernick, Richard J.

    2006-01-01

    A Hastelloy-X clad, sodium-potassium (NaK-78) cooled, moderated spectrum reactor using uranium zirconium hydride (UZrH) fuel based on the SNAP program reactors is a promising design for use in surface power systems. This paper presents a 98 kWth reactor for a power system the uses multiple Stirling engines to produce 25 kWe-net for 5 years. The design utilizes a pin type geometry containing UZrHx fuel clad with Hastelloy-X and NaK-78 flowing around the pins as coolant. A compelling feature of this design is its use of 49.9% enriched U, allowing it to be classified as a category III-D attractiveness and reducing facility costs relative to highly-enriched space reactor concepts. Presented below are both the design and an analysis of this reactor's criticality under various safety and operations scenarios.

  4. Current State of European Railway Fire Safety Research

    DOT National Transportation Integrated Search

    1985-06-01

    This report describes the recent fire safety research and practical fire experience of the major European railways. It includes a summary of the main causes and characteristics of railway vehicle fires, general approaches to the problem, and existing...

  5. Leveraging Safety Programs to Improve and Support Security Programs

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Leach, Janice; Snell, Mark K.; Pratt, R.

    2015-10-01

    There has been a long history of considering Safety, Security, and Safeguards (3S) as three functions of nuclear security design and operations that need to be properly and collectively integrated with operations. This paper specifically considers how safety programmes can be extended directly to benefit security as part of an integrated facility management programme. The discussion will draw on experiences implementing such a programme at Sandia National Laboratories’ Annular Research Reactor Facility. While the paper focuses on nuclear facilities, similar ideas could be used to support security programmes at other types of high-consequence facilities and transportation activities.

  6. Phenomena Important in Molten Salt Reactor Simulations

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Diamond, David J.; Brown, Nicholas R.; Denning, Richard

    The U.S. Nuclear Regulatory Commission (NRC) is preparing for the future licensing of advanced reactors that will be very different from current light water reactors. Part of the NRC preparation strategy is to identify the simulation tools that will be used for confirmatory safety analysis of normal operation and abnormal situations in those reactors. This report advances that strategy for reactors that will use molten salts (MSRs). This includes reactors with the fuel within the salt as well as reactors using solid fuel. Although both types are discussed in this report, the emphasis is on those reactors with liquid fuelmore » because of the perception that solid-fuel MSRs will be significantly easier to simulate. These liquid-fuel reactors include thermal and fast neutron spectrum alternatives. The specific designs discussed in the report are a subset of many designs being considered in the U.S. and elsewhere but they are considered the most likely to submit information to the NRC in the near future. The objective herein, is to understand the design of proposed molten salt reactors, how they will operate under normal or transient/accident conditions, and what will be the corresponding modeling needs of simulation tools that consider neutronics, heat transfer, fluid dynamics, and material composition changes in the molten salt. These tools will enable the NRC to eventually carry out confirmatory analyses that examine the validity and accuracy of applicant’s calculations and help determine the margin of safety in plant design.« less

  7. Design and analysis of a nuclear reactor core for innovative small light water reactors

    NASA Astrophysics Data System (ADS)

    Soldatov, Alexey I.

    In order to address the energy needs of developing countries and remote communities, Oregon State University has proposed the Multi-Application Small Light Water Reactor (MASLWR) design. In order to achieve five years of operation without refueling, use of 8% enriched fuel is necessary. This dissertation is focused on core design issues related with increased fuel enrichment (8.0%) and specific MASLWR operational conditions (such as lower operational pressure and temperature, and increased leakage due to small core). Neutron physics calculations are performed with the commercial nuclear industry tools CASMO-4 and SIMULATE-3, developed by Studsvik Scandpower Inc. The first set of results are generated from infinite lattice level calculations with CASMO-4, and focus on evaluation of the principal differences between standard PWR fuel and MASLWR fuel. Chapter 4-1 covers aspects of fuel isotopic composition changes with burnup, evaluation of kinetic parameters and reactivity coefficients. Chapter 4-2 discusses gadolinium self-shielding and shadowing effects, and subsequent impacts on power generation peaking and Reactor Control System shadowing. The second aspect of the research is dedicated to core design issues, such as reflector design (chapter 4-3), burnable absorber distribution and programmed fuel burnup and fuel use strategy (chapter 4-4). This section also includes discussion of the parameters important for safety and evaluation of Reactor Control System options for the proposed core design. An evaluation of the sensitivity of the proposed design to uncertainty in calculated parameters is presented in chapter 4-5. The results presented in this dissertation cover a new area of reactor design and operational parameters, and may be applicable to other small and large pressurized water reactor designs.

  8. 27. The top of a typical pile, F Reactor in ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    27. The top of a typical pile, F Reactor in February 1945 in this case, showing the vertical safety rods (VSRs) and the cables that support them. The rods could be dropped into the pile to effect a rapid shutdown. The four silvered-colored drums on the left contained boron solution and are part of the last ditch safety system. Should the VSRs channels become blocked by an occurrence such as an earthquake, the solution could be dumped into the VSR channels to help shut down the reactor. D-8334 - B Reactor, Richland, Benton County, WA

  9. Interim MELCOR Simulation of the Fukushima Daiichi Unit 2 Accident Reactor Core Isolation Cooling Operation

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Ross, Kyle W.; Gauntt, Randall O.; Cardoni, Jeffrey N.

    2013-11-01

    Data, a brief description of key boundary conditions, and results of Sandia National Laboratories’ ongoing MELCOR analysis of the Fukushima Unit 2 accident are given for the reactor core isolation cooling (RCIC) system. Important assumptions and related boundary conditions in the current analysis additional to or different than what was assumed/imposed in the work of SAND2012-6173 are identified. This work is for the U.S. Department of Energy’s Nuclear Energy University Programs fiscal year 2014 Reactor Safety Technologies Research and Development Program RC-7: RCIC Performance under Severe Accident Conditions.

  10. JHR Project: a future Material Testing Reactor working as an International user Facility: The key-role of instrumentation in support to the development of modern experimental capacity

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Bignan, G.; Gonnier, C.; Lyoussi, A.

    2015-07-01

    Research and development on fuel and material behaviour under irradiation is a key issue for sustainable nuclear energy in order to meet specific needs by keeping the best level of safety. These needs mainly deal with a constant improvement of performances and safety in order to optimize the fuel cycle and hence to reach nuclear energy sustainable objectives. A sustainable nuclear energy requires a high level of performances in order to meet specific needs such as: - Pursuing improvement of the performances and safety of present and coming water cooled reactor technologies. This will require a continuous R and Dmore » support following a long-term trend driven by the plant life management, safety demonstration, flexibility and economics improvement. Experimental irradiations of structure materials are necessary to anticipate these material behaviours and will contribute to their optimisation. - Upgrading continuously nuclear fuel technology in present and future nuclear power plants to achieve better performances and to optimise the fuel cycle keeping the best level of safety. Fuel evolution for generation II, III and III+ is a key stake requiring developments, qualification tests and safety experiments to ensure the competitiveness and safety: experimental tests exploring the full range of fuel behaviour determine fuel stability limits and safety margins, as a major input for the fuel reliability analysis. To perform such accurate and innovative progress and developments, specific and ad hoc instrumentation, irradiation devices, measurement methods are necessary to be set up inside or beside the material testing reactor (MTR) core. These experiments require beforehand in situ and on line sophisticated measurements to accurately determine different key parameters such as thermal and fast neutron fluxes and nuclear heating in order to precisely monitor and control the conducted assays. The new Material Testing Reactor JHR (Jules Horowitz Reactor) currently under

  11. Assuring safety without animal testing: Unilever's ongoing research programme to deliver novel ways to assure consumer safety.

    PubMed

    Westmoreland, Carl; Carmichael, Paul; Dent, Matt; Fentem, Julia; MacKay, Cameron; Maxwell, Gavin; Pease, Camilla; Reynolds, Fiona

    2010-01-01

    Assuring consumer safety without the generation of new animal data is currently a considerable challenge. However, through the application of new technologies and the further development of risk-based approaches for safety assessment, we remain confident it is ultimately achievable. For many complex, multi-organ consumer safety endpoints, the development, evaluation and application of new, non-animal approaches is hampered by a lack of biological understanding of the underlying mechanistic processes involved. The enormity of this scientific challenge should not be underestimated. To tackle this challenge a substantial research programme was initiated by Unilever in 2004 to critically evaluate the feasibility of a new conceptual approach based upon the following key components: 1.Developing new, exposure-driven risk assessment approaches. 2.Developing new biological (in vitro) and computer-based (in silico) predictive models. 3.Evaluating the applicability of new technologies for generating data (e.g. "omics", informatics) and for integrating new types of data (e.g. systems approaches) for risk-based safety assessment. Our research efforts are focussed in the priority areas of skin allergy, cancer and general toxicity (including inhaled toxicity). In all of these areas, a long-term investment is essential to increase the scientific understanding of the underlying biology and molecular mechanisms that we believe will ultimately form a sound basis for novel risk assessment approaches. Our research programme in these priority areas consists of in-house research as well as Unilever-sponsored academic research, involvement in EU-funded projects (e.g. Sens-it-iv, Carcinogenomics), participation in cross-industry collaborative research (e.g. Colipa, EPAA) and ongoing involvement with other scientific initiatives on non-animal approaches to risk assessment (e.g. UK NC3Rs, US "Human Toxicology Project" consortium).

  12. Development and implementation of participant safety plans for international research with stigmatised populations.

    PubMed

    Sugarman, Jeremy; Barnes, Mark; Rose, Scott; Dumchev, Kostyantyn; Sarasvita, Riza; Viet, Ha Tran; Zeziulin, Oleksandr; Susami, Hepa; Go, Vivian; Hoffman, Irving; Miller, William C

    2018-06-22

    People who inject drugs with high-risk sharing practices have high rates of HIV transmission and face barriers to HIV care. Interventions to overcome these barriers are needed; however, stigmatisation of drug use and HIV infection leads to safety concerns during the planning and conduct of research on such interventions. In preparing to address concerns about safety and wellbeing of participants in an international research study, HIV Prevention Trials Network 074, we developed participant safety plans (PSPs) at each site to supplement local research ethics committee oversight, community engagement, and usual clinical trial procedures. The PSPs were informed by systematic local legal and policy assessments, and interviews with key stakeholders. After PSP refinement and implementation, we assessed social impacts at each study visit to ensure continued safety. Throughout the study, five participants reported a negative social impact, with three resulting from study participation. Future research with stigmatised populations should consider using and assessing this approach to enhance safety and welfare. Copyright © 2018 Elsevier Ltd. All rights reserved.

  13. The role of simulation in mixed-methods research: a framework & application to patient safety.

    PubMed

    Guise, Jeanne-Marie; Hansen, Matthew; Lambert, William; O'Brien, Kerth

    2017-05-04

    Research in patient safety is an important area of health services research and is a national priority. It is challenging to investigate rare occurrences, explore potential causes, and account for the complex, dynamic context of healthcare - yet all are required in patient safety research. Simulation technologies have become widely accepted as education and clinical tools, but have yet to become a standard tool for research. We developed a framework for research that integrates accepted patient safety models with mixed-methods research approaches and describe the performance of the framework in a working example of a large National Institutes of Health (NIH)-funded R01 investigation. This worked example of a framework in action, identifies the strengths and limitations of qualitative and quantitative research approaches commonly used in health services research. Each approach builds essential layers of knowledge. We describe how the use of simulation ties these layers of knowledge together and adds new and unique dimensions of knowledge. A mixed-methods research approach that includes simulation provides a broad multi-dimensional approach to health services and patient safety research.

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

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1993-01-01

    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.

  15. Operating manual for the Bulk Shielding Reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Not Available

    1983-04-01

    The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxillary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supercedes all previous operating manuals for the BSR.

  16. Operating manual for the Bulk Shielding Reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Not Available

    1987-03-01

    The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxiliary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supersedes all previous operating manuals for the BSR.

  17. Predicting the effectiveness of road safety campaigns through alternative research designs.

    PubMed

    Adamos, Giannis; Nathanail, Eftihia

    2016-12-01

    A large number of road safety communication campaigns have been designed and implemented in the recent years; however their explicit impact on driving behavior and road accident rates has been estimated in a rather low proportion. Based on the findings of the evaluation of three road safety communication campaigns addressing the issues of drinking and driving, seat belt usage, and driving fatigue, this paper applies different types of research designs (i.e., experimental, quasi-experimental, and non-experimental designs), when estimating the effectiveness of road safety campaigns, implements a cross-design assessment, and conducts a cross-campaign evaluation. An integrated evaluation plan was developed, taking into account the structure of evaluation questions, the definition of measurable variables, the separation of the target audience into intervention (exposed to the campaign) and control (not exposed to the campaign) groups, the selection of alternative research designs, and the appropriate data collection methods and techniques. Evaluating the implementation of different research designs in estimating the effectiveness of road safety campaigns, results showed that the separate pre-post samples design demonstrated better predictability than other designs, especially in data obtained from the intervention group after the realization of the campaign. The more constructs that were added to the independent variables, the higher the values of the predictability were. The construct that most affects behavior is intention, whereas the rest of the constructs have a lower impact on behavior. This is particularly significant in the Health Belief Model (HBM). On the other hand, behavioral beliefs, normative beliefs, and descriptive norms, are significant parameters for predicting intention according to the Theory of Planned Behavior (TPB). The theoretical and applied implications of alternative research designs and their applicability in the evaluation of road safety

  18. Reactor physics teaching and research in the Swiss nuclear engineering master

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Chawla, R.; Paul Scherrer Inst., CH-5232 Villigen PSI

    Since 2008, a Master of Science program in Nuclear Engineering (NE) has been running in Switzerland, thanks to the combined efforts of the country's key players in nuclear teaching and research, viz. the Swiss Federal Inst.s of Technology at Lausanne (EPFL) and at Zurich (ETHZ), the Paul Scherrer Inst. (PSI) at Villigen and the Swiss Nuclear Utilities (Swissnuclear). The present paper, while outlining the academic program as a whole, lays emphasis on the reactor physics teaching and research training accorded to the students in the framework of the developed curriculum. (authors)

  19. The Safety Attitudes Questionnaire: psychometric properties, benchmarking data, and emerging research.

    PubMed

    Sexton, John B; Helmreich, Robert L; Neilands, Torsten B; Rowan, Kathy; Vella, Keryn; Boyden, James; Roberts, Peter R; Thomas, Eric J

    2006-04-03

    There is widespread interest in measuring healthcare provider attitudes about issues relevant to patient safety (often called safety climate or safety culture). Here we report the psychometric properties, establish benchmarking data, and discuss emerging areas of research with the University of Texas Safety Attitudes Questionnaire. Six cross-sectional surveys of health care providers (n = 10,843) in 203 clinical areas (including critical care units, operating rooms, inpatient settings, and ambulatory clinics) in three countries (USA, UK, New Zealand). Multilevel factor analyses yielded results at the clinical area level and the respondent nested within clinical area level. We report scale reliability, floor/ceiling effects, item factor loadings, inter-factor correlations, and percentage of respondents who agree with each item and scale. A six factor model of provider attitudes fit to the data at both the clinical area and respondent nested within clinical area levels. The factors were: Teamwork Climate, Safety Climate, Perceptions of Management, Job Satisfaction, Working Conditions, and Stress Recognition. Scale reliability was 0.9. Provider attitudes varied greatly both within and among organizations. Results are presented to allow benchmarking among organizations and emerging research is discussed. The Safety Attitudes Questionnaire demonstrated good psychometric properties. Healthcare organizations can use the survey to measure caregiver attitudes about six patient safety-related domains, to compare themselves with other organizations, to prompt interventions to improve safety attitudes and to measure the effectiveness of these interventions.

  20. The Safety Attitudes Questionnaire: psychometric properties, benchmarking data, and emerging research

    PubMed Central

    Sexton, John B; Helmreich, Robert L; Neilands, Torsten B; Rowan, Kathy; Vella, Keryn; Boyden, James; Roberts, Peter R; Thomas, Eric J

    2006-01-01

    Background There is widespread interest in measuring healthcare provider attitudes about issues relevant to patient safety (often called safety climate or safety culture). Here we report the psychometric properties, establish benchmarking data, and discuss emerging areas of research with the University of Texas Safety Attitudes Questionnaire. Methods Six cross-sectional surveys of health care providers (n = 10,843) in 203 clinical areas (including critical care units, operating rooms, inpatient settings, and ambulatory clinics) in three countries (USA, UK, New Zealand). Multilevel factor analyses yielded results at the clinical area level and the respondent nested within clinical area level. We report scale reliability, floor/ceiling effects, item factor loadings, inter-factor correlations, and percentage of respondents who agree with each item and scale. Results A six factor model of provider attitudes fit to the data at both the clinical area and respondent nested within clinical area levels. The factors were: Teamwork Climate, Safety Climate, Perceptions of Management, Job Satisfaction, Working Conditions, and Stress Recognition. Scale reliability was 0.9. Provider attitudes varied greatly both within and among organizations. Results are presented to allow benchmarking among organizations and emerging research is discussed. Conclusion The Safety Attitudes Questionnaire demonstrated good psychometric properties. Healthcare organizations can use the survey to measure caregiver attitudes about six patient safety-related domains, to compare themselves with other organizations, to prompt interventions to improve safety attitudes and to measure the effectiveness of these interventions. PMID:16584553

  1. Work organization research at the National Institute for Occupational Safety and Health.

    PubMed

    Rosenstock, L

    1997-01-01

    For 25 years, the National Institute for Occupational Safety and Health (NIOSH) has conducted and sponsored laboratory, field, and epidemiological studies that have helped define the role of work organization factors in occupational safety and health. Research has focused on the health effects of specific job conditions, occupational stressors in specific occupations, occupational difference in the incidence of stressors and stress-related disorders, and intervention strategies. NIOSH and the American Psychological Association have formalized the concept of occupational health psychology and developed a postdoctoral training program. The National Occupational Research Agenda recognizes organization of work as one of 21 national occupational safety and health research priority areas. Future research should focus on industries, occupations, and populations at special risk; the impact of work organization on overall health; the identification of healthy organization characteristics; and the development of intervention strategies.

  2. Unique educational opportunities at the Missouri University research reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Ketring, A.R.; Ross, F.K.; Spate, V.

    1997-12-01

    Since the Missouri University Research Reactor (MURR) went critical in 1966, it has been a center where students from many departments conduct their graduate research. In the past three decades, hundreds of graduate students from the MU departments of chemistry, physics, anthropology, nuclear engineering, etc., have received masters and doctoral degrees based on research using neutrons produced at MURR. More recently, the educational opportunities at MURR have been expanded to include undergraduate students and local high school students. Since 1989 MURR has participated in the National Science Foundation-funded Research Experience for Undergraduates (REU) program. As part of this program, undergraduatemore » students from universities and colleges throughout the United States come to MURR and get hands-on research experience during the summer. Another program, started in 1994 by the Nuclear Analysis Program at MURR, allows students from a local high school to conduct a neutron activation analysis (NAA) experiment. We also conduct tours of the center, where we describe the research and educational programs at MURR to groups of elementary school children, high school science teachers, state legislators, professional organizations, and many other groups.« less

  3. Manned space flight nuclear system safety. Volume 5: Nuclear System safety guidelines. Part 1: Space base nuclear safety

    NASA Technical Reports Server (NTRS)

    1972-01-01

    The design and operations guidelines and requirements developed in the study of space base nuclear system safety are presented. Guidelines and requirements are presented for the space base subsystems, nuclear hardware (reactor, isotope sources, dynamic generator equipment), experiments, interfacing vehicles, ground support systems, range safety and facilities. Cross indices and references are provided which relate guidelines to each other, and to substantiating data in other volumes. The guidelines are intended for the implementation of nuclear safety related design and operational considerations in future space programs.

  4. NEUTRONIC REACTOR FUEL ELEMENT

    DOEpatents

    Horning, W.A.; Lanning, D.D.; Donahue, D.J.

    1959-10-01

    A fuel slug for a reactor which acts as a safety device is described. The fuel slug is an aluminum tube with a foil lining the inside surface of the tube, the foil being fabricated of uranium in a lead matrix.

  5. Coupled reactor kinetics and heat transfer model for heat pipe cooled reactors

    NASA Astrophysics Data System (ADS)

    Wright, Steven A.; Houts, Michael

    2001-02-01

    Heat pipes are often proposed as cooling system components for small fission reactors. SAFE-300 and STAR-C are two reactor concepts that use heat pipes as an integral part of the cooling system. Heat pipes have been used in reactors to cool components within radiation tests (Deverall, 1973); however, no reactor has been built or tested that uses heat pipes solely as the primary cooling system. Heat pipe cooled reactors will likely require the development of a test reactor to determine the main differences in operational behavior from forced cooled reactors. The purpose of this paper is to describe the results of a systems code capable of modeling the coupling between the reactor kinetics and heat pipe controlled heat transport. Heat transport in heat pipe reactors is complex and highly system dependent. Nevertheless, in general terms it relies on heat flowing from the fuel pins through the heat pipe, to the heat exchanger, and then ultimately into the power conversion system and heat sink. A system model is described that is capable of modeling coupled reactor kinetics phenomena, heat transfer dynamics within the fuel pins, and the transient behavior of heat pipes (including the melting of the working fluid). This paper focuses primarily on the coupling effects caused by reactor feedback and compares the observations with forced cooled reactors. A number of reactor startup transients have been modeled, and issues such as power peaking, and power-to-flow mismatches, and loading transients were examined, including the possibility of heat flow from the heat exchanger back into the reactor. This system model is envisioned as a tool to be used for screening various heat pipe cooled reactor concepts, for designing and developing test facility requirements, for use in safety evaluations, and for developing test criteria for in-pile and out-of-pile test facilities. .

  6. Engaging policy makers in road safety research in Malaysia: a theoretical and contextual analysis.

    PubMed

    Tran, Nhan T; Hyder, Adnan A; Kulanthayan, Subramaniam; Singh, Suret; Umar, R S Radin

    2009-04-01

    Road traffic injuries (RTIs) are a growing public health problem that must be addressed through evidence-based interventions including policy-level changes such as the enactment of legislation to mandate specific behaviors and practices. Policy makers need to be engaged in road safety research to ensure that road safety policies are grounded in scientific evidence. This paper examines the strategies used to engage policy makers and other stakeholder groups and discusses the challenges that result from a multi-disciplinary, inter-sectoral collaboration. A framework for engaging policy makers in research was developed and applied to describe an example of collective road safety research in Malaysia. Key components of this framework include readiness, assessment, planning, implementation/evaluation, and policy development/sustainability. The case study of a collaborative intervention trial for the prevention of motorcycle crashes and deaths in Malaysia serves as a model for policy engagement by road safety and injury researchers. The analytic description of this research process in Malaysia demonstrates that the framework, through its five stages, can be used as a tool to guide the integration of needed research evidence into policy for road safety and injury prevention.

  7. Production and study of radionuclides at the research institute of atomic reactors (NIIAR)

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Karelin, E.A.; Gordeev, Y.N.; Filimonov, V.T.

    1995-01-01

    The main works of the Radionuclide Sources and Preparations Department (ORIP) of the Research Institute of Atomic Reactors (NIIAR) are summarized. The major activity of the Radionuclide Sources and Preparations Department (ORIP) is aimed at production of radioactive preparations of trans-plutonium elements (TPE) and also of lighter elements (from P to Ir), manufacture of ionizing radiation sources thereof, and scientific research to develop new technologies. One of the radionuclides that recently has received major attention is gadolinium-153. Photon sources based on it are used in densimeters for diagnostics of bone deseases. The procedure for separating gadolinium and europium, which ismore » currently used at the Research Institute of Atomic Reactors (NILAR), is based on europium cementation with the use of sodium amalgam. The method, though efficient, did not until recently permit an exhaustive removal of radioactive europium from {sup 153}Gd. The authors have thoroughly studied the separation process in semi-countercurrent mode, using citrate solutions. A special attention was given to the composition of europium complex species.« less

  8. MYRRHA: A multipurpose nuclear research facility

    NASA Astrophysics Data System (ADS)

    Baeten, P.; Schyns, M.; Fernandez, Rafaël; De Bruyn, Didier; Van den Eynde, Gert

    2014-12-01

    MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a multipurpose research facility currently being developed at SCK•CEN. MYRRHA is based on the ADS (Accelerator Driven System) concept where a proton accelerator, a spallation target and a subcritical reactor are coupled. MYRRHA will demonstrate the ADS full concept by coupling these three components at a reasonable power level to allow operation feedback. As a flexible irradiation facility, the MYRRHA research facility will be able to work in both critical as subcritical modes. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material developments for GEN IV and fusion reactors, and radioisotope production for medical and industrial applications. MYRRHA will be cooled by lead-bismuth eutectic and will play an important role in the development of the Pb-alloys technology needed for the LFR (Lead Fast Reactor) GEN IV concept. MYRRHA will also contribute to the study of partitioning and transmutation of high-level waste. Transmutation of minor actinides (MA) can be completed in an efficient way in fast neutron spectrum facilities, so both critical reactors and subcritical ADS are potential candidates as dedicated transmutation systems. However critical reactors heavily loaded with fuel containing large amounts of MA pose reactivity control problems, and thus safety problems. A subcritical ADS operates in a flexible and safe manner, even with a core loading containing a high amount of MA leading to a high transmutation rate. In this paper, the most recent developments in the design of the MYRRHA facility are presented.

  9. Neutronic reactor

    DOEpatents

    Wende, Charles W. J.

    1976-08-17

    A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield.

  10. Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

    NASA Astrophysics Data System (ADS)

    Cisneros, Anselmo Tomas, Jr.

    The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP

  11. Qualitative Research for Patient Safety Using ICTs: Methodological Considerations in the Technological Age.

    PubMed

    Yee, Kwang Chien; Wong, Ming Chao; Turner, Paul

    2017-01-01

    Considerable effort and resources have been dedicated to improving the quality and safety of patient care through health information systems, but there is still significant scope for improvement. One contributing factor to the lack of progress in patient safety improvement especially where technology has been deployed relates to an over-reliance on purely objective, quantitative, positivist research paradigms as the basis for generating and validating evidence of improvement. This paper argues the need for greater recognition and accommodation of evidence of improvement generated through more subjective, qualitative and pragmatic research paradigms to aid patient safety especially where technology is deployed. This paper discusses how acknowledging the role and value of more subjective ontologies and pragmatist epistemologies can support improvement science research. This paper illustrates some challenges and benefits from adopting qualitative research methods in patient safety improvement projects, particularly focusing challenges in the technological era. While adopting methods that can more readily capture, analyse and interpret direct user experiences, attitudes, insights and behaviours in their contextual settings, patient safety can be enhanced 'on the ground' and errors reduced and/or mitigated, challenges of using these methods with the younger "technologically-centred" healthcare professionals and patients needs to recognised.

  12. [Establish research model of post-marketing clinical safety evaluation for Chinese patent medicine].

    PubMed

    Zheng, Wen-ke; Liu, Zhi; Lei, Xiang; Tian, Ran; Zheng, Rui; Li, Nan; Ren, Jing-tian; Du, Xiao-xi; Shang, Hong-cai

    2015-09-01

    The safety of Chinese patent medicine has become a focus of social. It is necessary to carry out work on post-marketing clinical safety evaluation for Chinese patent medicine. However, there have no criterions to guide the related research, it is urgent to set up a model and method to guide the practice for related research. According to a series of clinical research, we put forward some views, which contained clear and definite the objective and content of clinical safety evaluation, the work flow should be determined, make a list of items for safety evaluation project, and put forward the three level classification of risk control. We set up a model of post-marketing clinical safety evaluation for Chinese patent medicine. Based this model, the list of items can be used for ranking medicine risks, and then take steps for different risks, aims to lower the app:ds:risksrisk level. At last, the medicine can be managed by five steps in sequence. The five steps are, collect risk signal, risk recognition, risk assessment, risk management, and aftereffect assessment. We hope to provide new ideas for the future research.

  13. On fundamental quality of fission chain reaction to oppose rapid runaways of nuclear reactors

    NASA Astrophysics Data System (ADS)

    Kulikov, G. G.; Shmelev, A. N.; Apse, V. A.; Kulikov, E. G.

    2017-01-01

    It has been shown that the in-hour equation characterizes the barriers and resistibility of fission chain reaction (FCR) against rapid runaways in nuclear reactors. Traditionally, nuclear reactors are characterized by the presence of barriers based on delayed and prompt neutrons. A new barrier based on the reflector neutrons that can occur when the fast reactor core is surrounded by a weakly absorbing neutron reflector with heavy atomic weight was proposed. It has been shown that the safety of this fast reactor is substantially improved, and considerable elongation of prompt neutron lifetime "devalues" the role of delayed neutron fraction as the maximum permissible reactivity for the reactor safety.

  14. On use of ZPR research reactors and associated instrumentation and measurement methods for reactor physics studies

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Chauvin, J.P.; Blaise, P.; Lyoussi, A.

    2015-07-01

    The French atomic and alternative energies -CEA- is strongly involved in research and development programs concerning the use of nuclear energy as a clean and reliable source of energy and consequently is working on the present and future generation of reactors on various topics such as ageing plant management, optimization of the plutonium stockpile, waste management and innovative systems exploration. Core physics studies are an essential part of this comprehensive R and D effort. In particular, the Zero Power Reactor (ZPR) of CEA: EOLE, MINERVE and MASURCA play an important role in the validation of neutron (as well photon) physicsmore » calculation tools (codes and nuclear data). The experimental programs defined in the CEA's ZPR facilities aim at improving the calculation routes by reducing the uncertainties of the experimental databases. They also provide accurate data on innovative systems in terms of new materials (moderating and decoupling materials) and new concepts (ADS, ABWR, new MTR (e.g. JHR), GENIV) involving new fuels, absorbers and coolant materials. Conducting such interesting experimental R and D programs is based on determining and measuring main parameters of phenomena of interest to qualify calculation tools and nuclear data 'libraries'. Determining these parameters relies on the use of numerous and different experimental techniques using specific and appropriate instrumentation and detection tools. Main ZPR experimental programs at CEA, their objectives and challenges will be presented and discussed. Future development and perspectives regarding ZPR reactors and associated programs will be also presented. (authors)« less

  15. Instrumental variable methods in comparative safety and effectiveness research.

    PubMed

    Brookhart, M Alan; Rassen, Jeremy A; Schneeweiss, Sebastian

    2010-06-01

    Instrumental variable (IV) methods have been proposed as a potential approach to the common problem of uncontrolled confounding in comparative studies of medical interventions, but IV methods are unfamiliar to many researchers. The goal of this article is to provide a non-technical, practical introduction to IV methods for comparative safety and effectiveness research. We outline the principles and basic assumptions necessary for valid IV estimation, discuss how to interpret the results of an IV study, provide a review of instruments that have been used in comparative effectiveness research, and suggest some minimal reporting standards for an IV analysis. Finally, we offer our perspective of the role of IV estimation vis-à-vis more traditional approaches based on statistical modeling of the exposure or outcome. We anticipate that IV methods will be often underpowered for drug safety studies of very rare outcomes, but may be potentially useful in studies of intended effects where uncontrolled confounding may be substantial.

  16. Metal fires and their implications for advanced reactors.

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Nowlen, Steven Patrick; Figueroa, Victor G.; Olivier, Tara Jean

    This report details the primary results of the Laboratory Directed Research and Development project (LDRD 08-0857) Metal Fires and Their Implications for Advance Reactors. Advanced reactors may employ liquid metal coolants, typically sodium, because of their many desirable qualities. This project addressed some of the significant challenges associated with the use of liquid metal coolants, primary among these being the extremely rapid oxidation (combustion) that occurs at the high operating temperatures in reactors. The project has identified a number of areas for which gaps existed in knowledge pertinent to reactor safety analyses. Experimental and analysis capabilities were developed in thesemore » areas to varying degrees. In conjunction with team participation in a DOE gap analysis panel, focus was on the oxidation of spilled sodium on thermally massive surfaces. These are spills onto surfaces that substantially cool the sodium during the oxidation process, and they are relevant because standard risk mitigation procedures seek to move spill environments into this regime through rapid draining of spilled sodium. While the spilled sodium is not quenched, the burning mode is different in that there is a transition to a smoldering mode that has not been comprehensively described previously. Prior work has described spilled sodium as a pool fire, but there is a crucial, experimentally-observed transition to a smoldering mode of oxidation. A series of experimental measurements have comprehensively described the thermal evolution of this type of sodium fire for the first time. A new physics-based model has been developed that also predicts the thermal evolution of this type of sodium fire for the first time. The model introduces smoldering oxidation through porous oxide layers to go beyond traditional pool fire analyses that have been carried out previously in order to predict experimentally observed trends. Combined, these developments add significantly to the

  17. Compendium of research and evaluations in traffic safety published

    DOT National Transportation Integrated Search

    1996-05-01

    The National Highway Traffic Safety Administration's (NHTSA's) Office of Program Development and Evaluation (OPDE) conducts research projects that investigate human attitudes, behaviors, and failures as they relate to motor vehicle crashes. OPDE focu...

  18. Flow characteristics of Korea multi-purpose research reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Heonil Kim; Hee Taek Chae; Byung Jin Jun

    1995-09-01

    The construction of Korea Multi-purpose Research Reactor (KMRR), a 30 MW{sub th} open-tank-in-pool type, is completed. Various thermal-hydraulic experiments have been conducted to verify the design characteristics of the KMRR. This paper describes the commissioning experiments to determine the flow distribution of KMRR core and the flow characteristics inside the chimney which stands on top of the core. The core flow is distributed to within {+-}6% of the average values, which is sufficiently flat in the sense that the design velocity in the fueled region is satisfied. The role of core bypass flow to confine the activated core coolant inmore » the chimney structure is confirmed.« less

  19. Benchmark Evaluation of Dounreay Prototype Fast Reactor Minor Actinide Depletion Measurements

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Hess, J. D.; Gauld, I. C.; Gulliford, J.

    2017-01-01

    Historic measurements of actinide samples in the Dounreay Prototype Fast Reactor (PFR) are of interest for modern nuclear data and simulation validation. Samples of various higher-actinide isotopes were irradiated for 492 effective full-power days and radiochemically assayed at Oak Ridge National Laboratory (ORNL) and Japan Atomic Energy Research Institute (JAERI). Limited data were available regarding the PFR irradiation; a six-group neutron spectra was available with some power history data to support a burnup depletion analysis validation study. Under the guidance of the Organisation for Economic Co-Operation and Development Nuclear Energy Agency (OECD NEA), the International Reactor Physics Experiment Evaluation Projectmore » (IRPhEP) and Spent Fuel Isotopic Composition (SFCOMPO) Project are collaborating to recover all measurement data pertaining to these measurements, including collaboration with the United Kingdom to obtain pertinent reactor physics design and operational history data. These activities will produce internationally peer-reviewed benchmark data to support validation of minor actinide cross section data and modern neutronic simulation of fast reactors with accompanying fuel cycle activities such as transportation, recycling, storage, and criticality safety.« less

  20. Collegiate Aviation Research and Education Solutions to Critical Safety Issues

    NASA Technical Reports Server (NTRS)

    Bowen, Brent (Editor)

    2002-01-01

    This Conference Proceedings is a collection of 6 abstracts and 3 papers presented April 19-20, 2001 in Denver, CO. The conference focus was "Best Practices and Benchmarking in Collegiate and Industry Programs". Topics covered include: satellite-based aviation navigation; weather safety training; human-behavior and aircraft maintenance issues; disaster preparedness; the collegiate aviation emergency response checklist; aviation safety research; and regulatory status of maintenance resource management.

  1. Operational performance of the three bean salad control algorithm on the ACRR (Annular Core Research Reactor)

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Ball, R.M.; Madaras, J.J.; Trowbridge, F.R. Jr.

    Experimental tests on the Annular Core Research Reactor have confirmed that the Three-Bean-Salad'' control algorithm based on the Pontryagin maximum principle can change the power of a nuclear reactor many decades with a very fast startup rate and minimal overshoot. The paper describes the results of simulations and operations up to 25 MW and 87 decades per minute. 3 refs., 4 figs., 1 tab.

  2. Automated startup of the MIT research reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Kwok, K.S.

    1992-01-01

    This summary describes the development, implementation, and testing of a generic method for performing automated startups of nuclear reactors described by space-independent kinetics under conditions of closed-loop digital control. The technique entails first obtaining a reliable estimate of the reactor's initial degree of subcriticality and then substituting that estimate into a model-based control law so as to permit a power increase from subcritical on a demanded trajectory. The estimation of subcriticality is accomplished by application of the perturbed reactivity method. The shutdown reactor is perturbed by the insertion of reactivity at a known rate. Observation of the resulting period permitsmore » determination of the initial degree of subcriticality. A major advantage to this method is that repeated estimates are obtained of the same quantity. Hence, statistical methods can be applied to improve the quality of the calculation.« less

  3. Overcoming recruitment challenges in construction safety intervention research.

    PubMed

    Kidd, Pamela; Parshall, Mark; Wojcik, Susan; Struttmann, Tim

    2004-03-01

    Recruiting workers in small construction companies and securing their participation in voluntary safety programs or safety research poses unique challenges. Worker turnover and worksite changes contribute to difficulties in locating and enrolling participants. Economic pressures and time demands potentially threaten ongoing participation. Six simulation exercises designed to reduce back and fall injuries in small construction companies were developed based on data from focus groups of workers and company owners. Working with a workers' compensation insurer, we had access to owner-operators of general, heavy, and special trade construction companies reporting less than $10,000 in payroll expenses. Recruitment methods included a participation incentive, mailed invitations followed by phone contacts, and follow-up reminders. Despite using recruitment methods recommended in the literature, participation rates were low over a 2-year intervention period. Because of these difficulties, factors affecting participation or nonparticipation became an additional research focus. Owners' perceptions of already having a good safety record and of the time demands of participation were the most commonly cited reasons for not participating. Literature on recruitment emphasizes processes and procedures under investigator control rather than understanding potential participants' judgments about the adequacy of their existing practices and the potential benefits of intervention participation relative to potential time and productivity trade-offs. Greater attention to such judgments may enhance recruitment and participation in under-studied and difficult to access populations. Copyright 2004 Wiley-Liss, Inc.

  4. Preliminary Development of Online Monitoring Acoustic Emission System for the Integrity of Research Reactor Components

    NASA Astrophysics Data System (ADS)

    Bakhri, S.; Sumarno, E.; Himawan, R.; Akbar, T. Y.; Subekti, M.; Sunaryo, G. R.

    2018-02-01

    Three research reactors owned by BATAN have been more than 25 years. Aging of (Structure, System and Component) SSC which is mainly related to mechanical causes become the most important issue for the sustainability and safety operation. Acoustic Emission (AE) is one of the appropriate and recommended methods by the IAEA for inspection as well as at the same time for the monitoring of mechanical SSC related. However, the advantages of AE method in detecting the acoustic emission both for the inspection and the online monitoring require a relatively complex measurement system including hardware software system for the signal detection and analysis purposes. Therefore, aim of this work was to develop an AE system based on an embedded system which capable for doing both the online monitoring and inspection of the research reactor’s integrity structure. An embedded system was selected due to the possibility to install the equipment on the field in extreme environmental condition with capability to store, analyses, and send the required information for further maintenance and operation. The research was done by designing the embedded system based on the Field Programmable Gate Array (FPGA) platform, because of their execution speed and system reconfigurable opportunities. The AE embedded system is then tested to identify the AE source location and AE characteristic under tensile material testing. The developed system successfully acquire the AE elastic waveform and determine the parameter-based analysis such as the amplitude, peak, duration, rise time, counts and the average frequency both for the source location test and the tensile test.

  5. Monochromatic neutron beam production at Brazilian nuclear research reactors

    NASA Astrophysics Data System (ADS)

    Stasiulevicius, Roberto; Rodrigues, Claudio; Parente, Carlos B. R.; Voi, Dante L.; Rogers, John D.

    2000-12-01

    Monochomatic beams of neutrons are obtained form a nuclear reactor polychromatic beam by the diffraction process, suing a single crystal energy selector. In Brazil, two nuclear research reactors, the swimming pool model IEA-R1 and the Argonaut type IEN-R1 have been used to carry out measurements with this technique. Neutron spectra have been measured using crystal spectrometers installed on the main beam lines of each reactor. The performance of conventional- artificial and natural selected crystals has been verified by the multipurpose neutron diffractometers installed at IEA-R1 and simple crystal spectrometer in operator at IEN- R1. A practical figure of merit formula was introduced to evaluate the performance and relative reflectivity of the selected planes of a single crystal. The total of 16 natural crystals were selected for use in the neutron monochromator, including a total of 24 families of planes. Twelve of these natural crystal types and respective best family of planes were measured directly with the multipurpose neutron diffractometers. The neutron spectrometer installed at IEN- R1 was used to confirm test results of the better specimens. The usually conventional-artificial crystal spacing distance range is limited to 3.4 angstrom. The interplane distance range has now been increased to approximately 10 angstrom by use of naturally occurring crystals. The neutron diffraction technique with conventional and natural crystals for energy selection and filtering can be utilized to obtain monochromatic sub and thermal neutrons with energies in the range of 0.001 to 10 eV. The thermal neutron is considered a good tool or probe for general applications in various fields, such as condensed matter, chemistry, biology, industrial applications and others.

  6. Soil bed reactor work of the Environmental Research Lab. of the University of Arizona in support of the research and development of Biosphere 2

    NASA Technical Reports Server (NTRS)

    Frye, Robert

    1990-01-01

    Research at the Environmental Research Lab in support of Biosphere 2 was both basic and applied in nature. One aspect of the applied research involved the use of biological reactors for the scrubbing of trace atmospheric organic contaminants. The research involved a quantitative study of the efficiency of operation of Soil Bed Reactors (SBR) and the optimal operating conditions for contaminant removal. The basic configuration of a SBR is that air is moved through a living soil that supports a population of plants. Upon exposure to the soil, contaminants are either passively adsorbed onto the surface of soil particles, chemically transformed in the soil to usable compounds that are taken up by the plants or microbes as a metabolic energy source and converted to CO2 and water.

  7. Decommissioning of German Research Reactors Under the Governance of the Federal Ministry of Education and Research - 12154

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Weigl, M.

    2012-07-01

    Since 1956, nuclear research and development (R and D) in Germany has been supported by the Federal Government. The goal was to help German industry to become competitive in all fields of nuclear technology. National research centers were established and demonstration plants were built. In the meantime, all these facilities were shut down and are now in a state of decommissioning and dismantling (D and D). Meanwhile, Germany is one of the leading countries in the world in the field of D and D. Two big demonstration plants, the Niederaichbach Nuclear Power Plant (KKN) a heavy-water cooled pressure tube reactormore » with carbon-dioxide cooling and the Karlstein Superheated Steam Reactor (HDR) a boiling light water reactor with a thermal power of 100 MW, are totally dismantled and 'green field' is reached. Another big project was finished in 2008. The Forschungs-Reaktor Juelich 1 (FRJ1), a research reactor with a thermal power of 10 MW was completely dismantled and in September 2008 an oak tree was planted on a green field at the site, where the FRJ1 was standing before. This is another example for German success in the field of D and D. Within these projects a lot of new solutions and innovative techniques were tested, which were developed at German universities and in small and medium sized companies mostly funded by the Federal Ministry of Education and Research (BMBF). Some examples are underwater-cutting technologies like plasma arc cutting and contact arc metal cutting. This clearly shows that research on the field of D and D is important for the future. Moreover, these research activities are important to save the know-how in nuclear engineering in Germany and will enable enterprises to compete on the increasing market of D and D services. The author assumes that an efficient decommissioning of nuclear installations will help stabilize the credibility of nuclear energy. Some critics of nuclear energy are insisting that a return to 'green field sites' is not

  8. Delivering safety

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Baldwin, N.D.; Spooner, K.G.; Walkden, P.

    2007-07-01

    In the United Kingdom there have been significant recent changes to the management of civil nuclear liabilities. With the formation in April 2005 of the Nuclear Decommissioning Authority (NDA), ownership of the civil nuclear licensed sites in the UK, including the Magnox Reactor Stations, passed to this new organisation. The NDAs mission is to seek acceleration of the nuclear clean up programme and deliver increased value for money and, consequently, are driving their contractors to seek more innovative ways of performing work. British Nuclear Group manages the UK Magnox stations under contract to the NDA. This paper summarises the approachmore » being taken within its Reactor Sites business to work with suppliers to enhance working arrangements at sites, improve the delivery of decommissioning programmes and deliver improvements in safety and environmental performance. The UK Magnox stations are 1. generation gas-graphite reactors, constructed in the 1950's and 1960's. Two stations are currently still operating, three are shut-down undergoing defueling and the other five are being decommissioned. Despite the distractions of industry restructuring, an uncompromising policy of demanding improved performance in conjunction with improved safety and environmental standards has been adopted. Over the past 5 years, this policy has resulted in step-changes in performance at Reactor Sites, with increased electrical output and accelerated defueling and decommissioning. The improvements in performance have been mirrored by improvements in safety (DACR of 0 at 5 sites); environmental standards (reductions in energy and water consumption, increased waste recycling) and the overall health of the workforce (20% reduction in sickness absence). These achievements have, in turn, been recognised by external bodies, resulting in several awards, including: the world's first ISRS and IERS level 10 awards (Sizewell, 2006), the NUMEX plant maintenance award (Bradwell, 2006), numerous

  9. Analyzing research trends on drug safety using topic modeling.

    PubMed

    Zou, Chen

    2018-06-01

    Published drug safety data has evolved in the past decade due to scientific and technological advances in the relevant research fields. Considering that a vast amount of scientific literature has been published in this area, it is not easy to identify the key information. Topic modeling has emerged as a powerful tool to extract meaningful information from a large volume of unstructured texts. Areas covered: We analyzed the titles and abstracts of 4347 articles in four journals dedicated to drug safety from 2007 to 2016. We applied Latent Dirichlet allocation (LDA) model to extract 50 main topics, and conducted trend analysis to explore the temporal popularity of these topics over years. Expert Opinion/Commentary: We found that 'benefit-risk assessment and communication', 'diabetes' and 'biologic therapy for autoimmune diseases' are the top 3 most published topics. The topics relevant to the use of electronic health records/observational data for safety surveillance are becoming increasingly popular over time. Meanwhile, there is a slight decrease in research on signal detection based on spontaneous reporting, although spontaneous reporting still plays an important role in benefit-risk assessment. The topics related to medical conditions and treatment showed highly dynamic patterns over time.

  10. Review of Transient Testing of Fast Reactor Fuels in the Transient REActor Test Facility (TREAT)

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Jensen, C.; Wachs, D.; Carmack, J.

    The restart of the Transient REActor Test (TREAT) facility provides a unique opportunity to engage the fast reactor fuels community to reinitiate in-pile experimental safety studies. Historically, the TREAT facility played a critical role in characterizing the behavior of both metal and oxide fast reactor fuels under off-normal conditions, irradiating hundreds of fuel pins to support fast reactor fuel development programs. The resulting test data has provided validation for a multitude of fuel performance and severe accident analysis computer codes. This paper will provide a review of the historical database of TREAT experiments including experiment design, instrumentation, test objectives, andmore » salient findings. Additionally, the paper will provide an introduction to the current and future experiment plans of the U.S. transient testing program at TREAT.« less

  11. The Experimental Breeder Reactor II seismic probabilistic risk assessment

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Roglans, J; Hill, D J

    1994-02-01

    The Experimental Breeder Reactor II (EBR-II) is a US Department of Energy (DOE) Category A research reactor located at Argonne National Laboratory (ANL)-West in Idaho. EBR-II is a 62.5 MW-thermal Liquid Metal Reactor (LMR) that started operation in 1964 and it is currently being used as a testbed in the Integral Fast Reactor (IFR) Program. ANL has completed a Level 1 Probabilistic Risk Assessment (PRA) for EBR-II. The Level 1 PRA for internal events and most external events was completed in June 1991. The seismic PRA for EBR-H has recently been completed. The EBR-II reactor building contains the reactor, themore » primary system, and the decay heat removal systems. The reactor vessel, which contains the core, and the primary system, consisting of two primary pumps and an intermediate heat exchanger, are immersed in the sodium-filled primary tank, which is suspended by six hangers from a beam support structure. Three systems or functions in EBR-II were identified as the most significant from the standpoint of risk of seismic-induced fuel damage: (1) the reactor shutdown system, (2) the structural integrity of the passive decay heat removal systems, and (3) the integrity of major structures, like the primary tank containing the reactor that could threaten both the reactivity control and decay heat removal functions. As part of the seismic PRA, efforts were concentrated in studying these three functions or systems. The passive safety response of EBR-II reactor -- both passive reactivity shutdown and passive decay heat removal, demonstrated in a series of tests in 1986 -- was explicitly accounted for in the seismic PRA as it had been included in the internal events assessment.« less

  12. Occupational health and safety surveillance and research using workers' compensation data.

    PubMed

    Utterback, David F; Schnorr, Teresa M; Silverstein, Barbara A; Spieler, Emily A; Leamon, Tom B; Amick, Benjamin C

    2012-02-01

    Examine uses of US workers' compensation (WC) data for occupational safety and health purposes. This article is a summary of the proceedings from an invitational workshop held in September 2009 to discuss the use of WC data for occupational safety and health prevention purposes. Workers' compensation data systems, although limited in many ways, contain information such as medical treatments, their costs and outcomes, and disability causes that are unavailable from national occupational surveillance sources. Despite their limitations, WC records are collected in a manner consistent with many occupational health and safety surveillance needs. Reports are available on the use of WC data for surveillance and research purposes such as estimating the frequency, magnitude, severity, and cost of compensated injuries. Inconsistencies in WC data can limit generalization of research results.

  13. Light-water-reactor safety research program. Quarterly progress report, July--September 1975

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Not Available

    1975-01-01

    Progress is summarized in the following research and development areas: (1) loss-of-coolant accident research; heat transfer and fluid dynamics; (2) transient fuel response and fission-product release; and (3) mechanical properties of Zircaloy containing oxygen. Also included is an appendix on Kinetics of Fission Gas and Volatile Fission-product Behavior under Transient Conditions in LWR Fuel.

  14. The shutdown reactor: Optimizing spent fuel storage cost

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Pennington, C.W.

    1995-12-31

    Several studies have indicated that the most prudent way to store fuel at a shutdown reactor site safely and economically is through the use of a dry storage facility licensed under 10CFR72. While such storage is certainly safe, is it true that the dry ISFSI represents the safest and most economical approach for the utility? While no one is really able to answer that question definitely, as yet, Holtec has studied this issue for some time and believes that both an economic and safety case can be made for an optimization strategy that calls for the use of both wetmore » and dry ISFSI storage of spent fuel at some plants. For the sake of brevity, this paper summarizes some of Holtec`s findings with respect to the economics of maintaining some fuel in wet storage at a shutdown reactor. The safety issue, or more importantly the perception of safety of spent fuel in wet storage, still varies too much with the eye of the beholder, and until a more rigorous presentation of safety analyses can be made in a regulatory setting, it is not practically useful to argue about how many angels can sit on the head of a safety-related pin. Holtec is prepared to present such analyses, but this does not appear to be the proper venue. Thus, this paper simply looks at certain economic elements of a wet ISFSI at a shutdown reactor to make a prima facie case that wet storage has some attractiveness at a shutdown reactor and should not be rejected out of hand. Indeed, an optimization study at certain plants may well show the economic vitality of keeping some fuel in the pool and converting the NRC licensing coverage from 10CFR50 to 10CFR72. If the economics look attractive, then the safety issue may be confronted with a compelling interest.« less

  15. Disposition of fuel elements from the Aberdeen and Sandia pulse reactor (SPR-II) assemblies

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Mckerley, Bill; Bustamante, Jacqueline M; Costa, David A

    2010-01-01

    We describe the disposition of fuel from the Aberdeen (APR) and the Sandia Pulse Reactors (SPR-II) which were used to provide intense neutron bursts for radiation effects testing. The enriched Uranium - 10% Molybdenum fuel from these reactors was shipped to the Los Alamos National Laboratory (LANL) for size reduction prior to shipment to the Savannah River Site (SRS) for final disposition in the H Canyon facility. The Shipper/Receiver Agreements (SRA), intra-DOE interfaces, criticality safety evaluations, safety and quality requirements and key materials management issues required for the successful completion of this project will be presented. This work is inmore » support of the DOE Consolidation and Disposition program. Sandia National Laboratories (SNL) has operated pulse nuclear reactor research facilities for the Department of Energy since 1961. The Sandia Pulse Reactor (SPR-II) was a bare metal Godiva-type reactor. The reactor facilities have been used for research and development of nuclear and non-nuclear weapon systems, advanced nuclear reactors, reactor safety, simulation sources and energy related programs. The SPR-II was a fast burst reactor, designed and constructed by SNL that became operational in 1967. The SPR-ll core was a solid-metal fuel enriched to 93% {sup 235}U. The uranium was alloyed with 10 weight percent molybdenum to ensure the phase stabilization of the fuel. The core consisted of six fuel plates divided into two assemblies of three plates each. Figure 1 shows a cutaway diagram of the SPR-II Reactor with its decoupling shroud. NNSA charged Sandia with removing its category 1 and 2 special nuclear material by the end of 2008. The main impetus for this activity was based on NNSA Administrator Tom D'Agostino's six focus areas to reenergize NNSA's nuclear material consolidation and disposition efforts. For example, the removal of SPR-II from SNL to DAF was part of this undertaking. This project was in support of NNSA's efforts to

  16. Human Hallucinogen Research: Guidelines for Safety

    PubMed Central

    Johnson, Matthew W.; Richards, William A.; Griffiths, Roland R.

    2010-01-01

    There has recently been a renewal of human research with classical hallucinogens (psychedelics). This paper first briefly discusses the unique history of human hallucinogen research, and then reviews the risks of hallucinogen administration and safeguards for minimizing these risks. Although hallucinogens are relatively safe physiologically and are not considered drugs of dependence, their administration involves unique psychological risks. The most likely risk is overwhelming distress during drug action (“bad trip”), which could lead to potentially dangerous behavior such as leaving the study site. Less common are prolonged psychoses triggered by hallucinogens. Safeguards against these risks include the exclusion of volunteers with personal or family history of psychotic disorders or other severe psychiatric disorders, establishing trust and rapport between session monitors and volunteer before the session, careful volunteer preparation, a safe physical session environment, and interpersonal support from at least two study monitors during the session. Investigators should probe for the relatively rare hallucinogen persisting perception disorder in follow up contact. Persisting adverse reactions are rare when research is conducted along these guidelines. Incautious research may jeopardize participant safety and future research. However, carefully conducted research may inform the treatment of psychiatric disorders, and may lead to advances in basic science. PMID:18593734

  17. Inherent Safety Characteristics of Advanced Fast Reactors

    NASA Astrophysics Data System (ADS)

    Bochkarev, A. S.; Korsun, A. S.; Kharitonov, V. S.; Alekseev, P. N.

    2017-01-01

    The study presents SFR transient performance for ULOF events initiated by pump trip and pump seizure with simultaneous failure of all shutdown systems in both cases. The most severe cases leading to the pin cladding rupture and possible sodium boiling are demonstrated. The impact of various features on SFR inherent safety performance for ULOF events was analysed. The decrease in hydraulic resistance of primary loop and increase in primary pump coast down time were investigated. Performing analysis resulted in a set of recommendations to varying parameters for the purpose of enhancing the inherent safety performance of SFR. In order to prevent the safety barrier rupture for ULOF events the set of thermal hydraulic criteria defining the ULOF transient processes dynamics and requirements to these criteria were recommended based on achieved results: primary sodium flow dip under the natural circulation asymptotic level and natural circulation rise time.

  18. Passive cooling system for top entry liquid metal cooled nuclear reactors

    DOEpatents

    Boardman, Charles E.; Hunsbedt, Anstein; Hui, Marvin M.

    1992-01-01

    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.

  19. Cultural safety, diversity and the servicer user and carer movement in mental health research.

    PubMed

    Cox, Leonie G; Simpson, Alan

    2015-12-01

    This study will be of interest to anyone concerned with a critical appraisal of mental health service users' and carers' participation in research collaboration and with the potential of the postcolonial paradigm of cultural safety to contribute to the service user research (SUR) movement. The history and nature of the mental health field and its relationship to colonial processes provokes a consideration of whether cultural safety could focus attention on diversity, power imbalance, cultural dominance and structural inequality, identified as barriers and tensions in SUR. We consider these issues in the context of state-driven approaches towards SUR in planning and evaluation and the concurrent rise of the SUR movement in the UK and Australia, societies with an intimate involvement in processes of colonisation. We consider the principles and motivations underlying cultural safety and SUR in the context of the policy agenda informing SUR. We conclude that while both cultural safety and SUR are underpinned by social constructionism constituting similarities in principles and intent, cultural safety has additional dimensions. Hence, we call on researchers to use the explicitly political and self-reflective process of cultural safety to think about and address issues of diversity, power and social justice in research collaboration. © 2015 John Wiley & Sons Ltd.

  20. Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core.

    PubMed

    Lashkari, A; Khalafi, H; Kazeminejad, H

    2013-05-01

    In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change.

  1. Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

    PubMed Central

    Lashkari, A.; Khalafi, H.; Kazeminejad, H.

    2013-01-01

    In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change. PMID:24976672

  2. Development of New Transportation/Storage Cask System for Use by DOE Russian Research Reactor Fuel Return Program

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Michael Tyacke; Frantisek Svitak; Jiri Rychecky

    2010-04-01

    The United States, the Russian Federation, and the International Atomic Energy Agency (IAEA) have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program. The purpose of this program is to return Soviet or Russian supplied high-enriched uranium (HEU) fuel currently stored at Russian-designed research reactors throughout the world to Russia. To accommodate transport of the HEU spent nuclear fuel (SNF), a new large-capacity transport/storage cask system was specially designed for handling and operations under the unique conditions for these research reactor facilities. This new cask system is named the ŠKODA VPVR/M cask. The design,more » licensing, testing, and delivery of this new cask system are the results of a significant international cooperative effort by several countries and involved numerous private and governmental organizations. This paper contains the following sections: (1) Introduction/Background; (2) VPVR/M Cask Description; (3) Ancillary Equipment, (4) Cask Licensing; (5) Cask Demonstration and Operations; (6) IAEA Procurement, Quality Assurance Inspections, Fabrication, and Delivery; and, (7) Summary and Conclusions.« less

  3. 100 years of occupational safety research: From basic protections and work analysis to a multilevel view of workplace safety and risk.

    PubMed

    Hofmann, David A; Burke, Michael J; Zohar, Dov

    2017-03-01

    Starting with initiatives dating back to the mid-1800s, we provide a high-level review of the key trends and developments in the application of applied psychology to the field of occupational safety. Factory laws, basic worker compensation, and research on accident proneness comprised much of the early work. Thus, early research and practice very much focused on the individual worker, the design of their work, and their basic protection. Gradually and over time, the focus began to navigate further into the organizational context. One of the early efforts to broaden beyond the individual worker was a significant focus on safety-related training during the middle of the 20th century. Toward the latter years of the 20th century and continuing the move from the individual worker to the broader organizational context, there was a significant increase in leadership and organizational climate (safety climate) research. Ultimately, this resulted in the development of a multilevel model of safety culture/climate. After discussing these trends, we identify key conclusions and opportunities for future research. (PsycINFO Database Record (c) 2017 APA, all rights reserved).

  4. Automatic reactor control system for transient operation

    NASA Astrophysics Data System (ADS)

    Lipinski, Walter C.; Bhattacharyya, Samit K.; Hanan, Nelson A.

    Various programmatic considerations have delayed the upgrading of the TREAT reactor and the performance of the control system is not yet experimentally verified. The current schedule calls for the upgrading activities to occur last in the calendar year 1987. Detailed simulation results, coupled with earlier validation of individual components of the control strategy in TREAT, verify the performance of the algorithms. The control system operates within the safety envelope provided by a protection system designed to ensure reactor safety under conditions of spurious reactivity additions. The approach should be directly applicable to MMW systems, with appropriate accounting of temperature rate limitations of key components and of the inertia of the secondary system components.

  5. Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

    NASA Astrophysics Data System (ADS)

    Ilham, Muhammad; Su'ud, Zaki

    2017-01-01

    Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

  6. Research on pressure control of pressurizer in pressurized water reactor nuclear power plant

    NASA Astrophysics Data System (ADS)

    Dai, Ling; Yang, Xuhong; Liu, Gang; Ye, Jianhua; Qian, Hong; Xue, Yang

    2010-07-01

    Pressurizer is one of the most important components in the nuclear reactor system. Its function is to keep the pressure of the primary circuit. It can prevent shutdown of the system from the reactor accident under the normal transient state while keeping the setting value in the normal run-time. This paper is mainly research on the pressure system which is running in the Daya Bay Nuclear Power Plant. A conventional PID controller and a fuzzy controller are designed through analyzing the dynamic characteristics and calculating the transfer function. Then a fuzzy PID controller is designed by analyzing the results of two controllers. The fuzzy PID controller achieves the optimal control system finally.

  7. A PC-based high temperature gas reactor simulator for Indonesian conceptual HTR reactor basic training

    NASA Astrophysics Data System (ADS)

    Syarip; Po, L. C. C.

    2018-05-01

    In planning for nuclear power plant construction in Indonesia, helium cooled high temperature reactor (HTR) is favorable for not relying upon water supply that might be interrupted by earthquake. In order to train its personnel, BATAN has cooperated with Micro-Simulation Technology of USA to develop a 200 MWt PC-based simulation model PCTRAN/HTR. It operates in Win10 environment with graphic user interface (GUI). Normal operation of startup, power maneuvering, shutdown and accidents including pipe breaks and complete loss of AC power have been conducted. A sample case of safety analysis simulation to demonstrate the inherent safety features of HTR was done for helium pipe break malfunction scenario. The analysis was done for the variation of primary coolant pipe break i.e. from 0,1% - 0,5 % and 1% - 10 % helium gas leakages, while the reactor was operated at the maximum constant power of 10 MWt. The result shows that the highest temperature of HTR fuel centerline and coolant were 1150 °C and 1296 °C respectively. With 10 kg/s of helium flow in the reactor core, the thermal power will back to the startup position after 1287 s of helium pipe break malfunction.

  8. Station Blackout Analysis of HTGR-Type Experimental Power Reactor

    NASA Astrophysics Data System (ADS)

    Syarip; Zuhdi, Aliq; Falah, Sabilul

    2018-01-01

    The National Nuclear Energy Agency of Indonesia has decided to build an experimental power reactor of high-temperature gas-cooled reactor (HTGR) type located at Puspiptek Complex. The purpose of this project is to demonstrate a small modular nuclear power plant that can be operated safely. One of the reactor safety characteristics is the reliability of the reactor to the station blackout (SBO) event. The event was observed due to relatively high disturbance frequency of electricity network in Indonesia. The PCTRAN-HTR functional simulator code was used to observe fuel and coolant temperature, and coolant pressure during the SBO event. The reactor simulated at 10 MW for 7200 s then the SBO occurred for 1-3 minutes. The analysis result shows that the reactor power decreases automatically as the temperature increase during SBO accident without operator’s active action. The fuel temperature increased by 36.57 °C every minute during SBO and the power decreased by 0.069 MW every °C fuel temperature rise at the condition of anticipated transient without reactor scram. Whilst, the maximum coolant (helium) temperature and pressure are 1004 °C and 9.2 MPa respectively. The maximum fuel temperature is 1282 °C, this value still far below the fuel temperature limiting condition i.e. 1600 °C, its mean that the HTGR has a very good inherent safety system.

  9. Calculation of the temperature in the container unit with a modified design for the production of {sup 99}Mo at the VVR-Ts research reactor facility (IVV.10M)

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Kazantsev, A. A., E-mail: kazantsevanatoly@gmail.com; Sergeev, V. V.; Kochnov, O. Yu.

    The temperature regime is calculated for two different designs of containers with uranium-bearing material for the upgraded VVR-Ts research reactor facility (IVV.10M). The containers are to be used in the production of {sup 99}Mo. It is demonstrated that the modification of the container design leads to a considerable temperature reduction and an increase in the near-wall boiling margin and allows one to raise the amount of material loaded into the container. The calculations were conducted using the international thermohydraulic contour code TRAC intended to analyze the technical safety of water-cooled nuclear power units.

  10. CESAR5.3: Isotopic depletion for Research and Testing Reactor decommissioning

    NASA Astrophysics Data System (ADS)

    Ritter, Guillaume; Eschbach, Romain; Girieud, Richard; Soulard, Maxime

    2018-05-01

    , CESAR includes a portable Graphical User Interface which can be broadly deployed in R&D or industrial facilities. Aging facilities currently face decommissioning and dismantling issues. This way to the end of the nuclear fuel cycle requires a careful assessment of source terms in the fuel, core structures and all parts of a facility that must be disposed of with "industrial nuclear" constraints. In that perspective, several CESAR cross section libraries were constructed for early CEA Research and Testing Reactors (RTR's). The aim of this paper is to describe how CESAR operates and how it can be used to help these facilities care for waste disposal, nuclear materials transport or basic safety cases. The test case will be based on the PHEBUS Facility located at CEA - Cadarache.

  11. Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Peterson, Per; Greenspan, Ehud

    2015-02-09

    This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designsmore » are used, the power density of salt- cooled reactors is limited to 10 MW/m 3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m 3. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X-PREX facility uses

  12. Collegiate Aviation Research and Education Solutions to Critical Safety Issues. UNO Aviation Monograph Series. UNOAI Report.

    ERIC Educational Resources Information Center

    Bowen, Brent, Ed.

    This document contains four papers concerning collegiate aviation research and education solutions to critical safety issues. "Panel Proposal Titled Collegiate Aviation Research and Education Solutions to Critical Safety Issues for the Tim Forte Collegiate Aviation Safety Symposium" (Brent Bowen) presents proposals for panels on the…

  13. KINETICS OF LOW SOURCE REACTOR STARTUPS. PART I

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Hurwitz, H. Jr.; MacMillan, D.B.; Smith, J.H.

    1962-06-01

    Statistical fluctuntions of neutron populations in reactors are analyzed by means of an approximate theoretical model. Development of the model is given in detail; also included are extensive numerical results derived from its application to systems with time-dependent reactivity, namely, a reactor during start-up. The special relationships of fluctuations to safety considerations are discussed. (auth)

  14. Fall Prevention Research and Practice: A Total Worker Safety Approach

    PubMed Central

    HSIAO, Hongwei

    2014-01-01

    Slips, trips, and falls (STF) represent a serious hazard to workers and occupants in many industries, homes, and communities. Often, the cause of a STF incident is multifactorial, encompassing human, environmental, and task risk factors. A STF-related disability can greatly diminish the occupational capability and quality of life of individuals in both the workplace and the home. Countering STF hazards and risks both on and off the job and on all aspects of control measures is a “total worker safety” matter, a challenging yet tangible undertaking. As the federal organization responsible for conducting research for the prevention of work-related injuries in the United States, the National Institute for Occupational Safety and Health (NIOSH) has been conducting research on STF controls for some decades. Many NIOSH research outcomes have been utilized for STF prevention in workplaces, with potential for prevention in homes as well. This paper summarizes the concept of total worker safety for STF control, NIOSH priority research goals, major activities, and accomplishments, and some emerging issues on STF. The strategic planning process for the NIOSH research goals and some identified research focuses are applicable to the development and implementation of global STF research goals. PMID:25345424

  15. Delayed Gamma Measurements in Different Nuclear Research Reactors Bringing Out the Importance of the Delayed Contribution in Gamma Flux Calculations

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Fourmentel, D.; Radulovic, V.; Barbot, L.

    Neutron and gamma flux levels are key parameters in nuclear research reactors. In Material Testing Reactors, such as the future Jules Horowitz Reactor, under construction at the French Alternative Energies and Atomic Energy Commission (CEA Cadarache, France), the expected gamma flux levels are very high (nuclear heating is of the order of 20 W/g at 100 MWth). As gamma rays deposit their energy in the reactor structures and structural materials it is important to take them into account when designing irradiation devices. There are only a few sensors which allow measurements of the nuclear heating ; a recent development atmore » the CEA Cadarache allows measurements of the gamma flux using a miniature ionization chamber (MIC). The measured MIC response is often compared with calculation using modern Monte Carlo (MC) neutron and photon transport codes, such as TRIPOLI-4 and MCNP6. In these calculations only the production of prompt gamma rays in the reactor is usually modelled thus neglecting the delayed gamma rays. Hence calculations and measurements are usually in better accordance for the neutron flux than for the gamma flux. In this paper we study the contribution of delayed gamma rays to the total MIC signal in order to estimate the systematic error in gamma flux MC calculations. In order to experimentally determine the delayed gamma flux contributions to the MIC response, we performed gamma flux measurements with CEA developed MIC at three different research reactors: the OSIRIS reactor (MTR - 70 MWth at CEA Saclay, France), the TRIGA MARK II reactor (TRIGA - 250 kWth at the Jozef Stefan Institute, Slovenia) and the MARIA reactor (MTR - 30 MWth at the National Center for Nuclear Research, Poland). In order to experimentally assess the delayed gamma flux contribution to the total gamma flux, several reactor shut down (scram) experiments were performed specifically for the purpose of the measurements. Results show that on average about 30 % of the MIC signal is

  16. Assessment of contributions to patient safety knowledge by the Agency for Healthcare Research and Quality-funded patient safety projects.

    PubMed

    Sorbero, Melony E S; Ricci, Karen A; Lovejoy, Susan; Haviland, Amelia M; Smith, Linda; Bradley, Lily A; Hiatt, Liisa; Farley, Donna O

    2009-04-01

    To characterize the activities of projects funded in Agency for Healthcare Research and Quality (AHRQ)'s patient safety portfolio and assess their aggregate potential to contribute to knowledge development. Information abstracted from proposals for projects funded in AHRQ's patient safety portfolio, information on safety practices from the AHRQ Evidence Report on Patient Safety Practices, and products produced by the projects. This represented one part of the process evaluation conducted as part of a longitudinal evaluation based on the Context–Input–Process–Product model. The 234 projects funded through AHRQ's patient safety portfolio examined a wide variety of patient safety issues and extended their work beyond the hospital setting to less studied parts of the health care system. Many of the projects implemented and tested practices for which the patient safety evidence report identified a need for additional evidence. The funded projects also generated a substantial body of new patient safety knowledge through a growing number of journal articles and other products. The projects funded in AHRQ's patient safety portfolio have the potential to make substantial contributions to the knowledge base on patient safety. The full value of this new knowledge remains to be confirmed through the synthesis of results

  17. Research Needs in Fire Safety for the Human Exploration and Utilization of Space

    NASA Technical Reports Server (NTRS)

    Ruff, Gary A.

    2003-01-01

    This viewgraph presentation provides information on developments in spacecraft fire safety research. The presentation includes an overview of the previous Spacecraft Fire Safety Workshop, from 1986, and the influences since then of bioastronautics on combustion science and fire safety. The presentation then gives of overview of the current conference, stating goals and giving a schedule.

  18. Manned space flight nuclear system safety. Volume 1: base nuclear system safety

    NASA Technical Reports Server (NTRS)

    1972-01-01

    The mission and terrestrial nuclear safety aspects of future long duration manned space missions in low earth orbit are discussed. Nuclear hazards of a typical low earth orbit Space Base mission (from natural sources and on-board nuclear hardware) have been identified and evaluated. Some of the principal nuclear safety design and procedural considerations involved in launch, orbital, and end of mission operations are presented. Areas of investigation include radiation interactions with the crew, subsystems, facilities, experiments, film, interfacing vehicles, nuclear hardware and the terrestrial populace. Results of the analysis indicate: (1) the natural space environment can be the dominant radiation source in a low earth orbit where reactors are effectively shielded, (2) with implementation of safety guidelines the reactor can present a low risk to the crew, support personnel, the terrestrial populace, flight hardware and the mission, (3) ten year missions are feasible without exceeding integrated radiation limits assigned to flight hardware, and (4) crew stay-times up to one year are feasible without storm shelter provisions.

  19. Comparative study between single core model and detail core model of CFD modelling on reactor core cooling behaviour

    NASA Astrophysics Data System (ADS)

    Darmawan, R.

    2018-01-01

    Nuclear power industry is facing uncertainties since the occurrence of the unfortunate accident at Fukushima Daiichi Nuclear Power Plant. The issue of nuclear power plant safety becomes the major hindrance in the planning of nuclear power program for new build countries. Thus, the understanding of the behaviour of reactor system is very important to ensure the continuous development and improvement on reactor safety. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last four decades 1D system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. Recently, 2D and 3D system level codes such as CFD are being explored. This paper discusses a comparative study on two different approaches of CFD modelling on reactor core cooling behaviour.

  20. Small reactor power systems for manned planetary surface bases

    NASA Technical Reports Server (NTRS)

    Bloomfield, Harvey S.

    1987-01-01

    A preliminary feasibility study of the potential application of small nuclear reactor space power systems to manned planetary surface base missions was conducted. The purpose of the study was to identify and assess the technology, performance, and safety issues associated with integration of reactor power systems with an evolutionary manned planetary surface exploration scenario. The requirements and characteristics of a variety of human-rated modular reactor power system configurations selected for a range of power levels from 25 kWe to hundreds of kilowatts is described. Trade-off analyses for reactor power systems utilizing both man-made and indigenous shielding materials are provided to examine performance, installation and operational safety feasibility issues. The results of this study have confirmed the preliminary feasibility of a wide variety of small reactor power plant configurations for growth oriented manned planetary surface exploration missions. The capability for power level growth with increasing manned presence, while maintaining safe radiation levels, was favorably assessed for nominal 25 to 100 kWe modular configurations. No feasibility limitations or technical barriers were identified and the use of both distance and indigenous planetary soil material for human rated radiation shielding were shown to be viable and attractive options.

  1. Westinghouse Small Modular Reactor nuclear steam supply system design

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Memmott, M. J.; Harkness, A. W.; Van Wyk, J.

    2012-07-01

    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 andmore » 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

  2. Fusion Safety Program annual report, fiscal year 1994

    NASA Astrophysics Data System (ADS)

    Longhurst, Glen R.; Cadwallader, Lee C.; Dolan, Thomas J.; Herring, J. Stephen; McCarthy, Kathryn A.; Merrill, Brad J.; Motloch, Chester C.; Petti, David A.

    1995-03-01

    This report summarizes the major activities of the Fusion Safety Program in fiscal year 1994. The Idaho National Engineering Laboratory (INEL) is the designated lead laboratory and Lockheed Idaho Technologies Company is the prime contractor for this program. The Fusion Safety Program was initiated in 1979. Activities are conducted at the INEL, at other DOE laboratories, and at other institutions, including the University of Wisconsin. The technical areas covered in this report include tritium safety, beryllium safety, chemical reactions and activation product release, safety aspects of fusion magnet systems, plasma disruptions, risk assessment failure rate data base development, and thermalhydraulics code development and their application to fusion safety issues. Much of this work has been done in support of the International Thermonuclear Experimental Reactor (ITER). Also included in the report are summaries of the safety and environmental studies performed by the Fusion Safety Program for the Tokamak Physics Experiment and the Tokamak Fusion Test Reactor and of the technical support for commercial fusion facility conceptual design studies. A major activity this year has been work to develop a DOE Technical Standard for the safety of fusion test facilities.

  3. Nuclear Fuel Traces Definition in Storage Ponds of Research VVR-2 and OR Reactors in NRC 'Kurchatov Institute'

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Stepanov, Alexey; Simirskii, Iurii; Stepanov, Vyacheslav

    2015-07-01

    The Gas Plant complex is the experimental base of the Institute of Nuclear Reactors, which is part of the Kurchatov Institute. In 1954 the commissioning of the first Soviet water-cooled water-moderated research reactor VVR-2 on enriched uranium, and until 1983 the complex operated two research water-cooled water-moderated reactors 3 MW (VVR-2) and 300 kW (OR) capacity, which were dismantled in connection with the overall upgrades of the complex. The complex has three storage ponds in the reactor building. They are sub-surface vessels filled with water (the volume of water in each is about 6 m{sup 3}). In 2007-2013 the spentmore » nuclear fuel from storages was removed for processing to 'Mayk'. Survey of Storage Ponds by Underwater Collimated Spectrometric System shows a considerable layer of slime on the bottom of ponds and traces of spent nuclear fuel in one of the storage. For determination qualitative and the quantitative composition of radionuclide we made complex α-, β-, γ- spectrometric research of water and bottom slimes from Gas Plant complex storage ponds. We found the spent nuclear fuel in water and bottom slime in all storage ponds. Specific activity of radionuclides in the bottom slime exceeded specific activity of radionuclides in the ponds water and was closed to levels of high radioactive waste. Analysis of the obtained data and data from earlier investigation of reactor MR storage ponds showed distinctions of specific activity of uranium and plutonium radionuclides. (authors)« less

  4. Solution of heat removal from nuclear reactors by natural convection

    NASA Astrophysics Data System (ADS)

    Zitek, Pavel; Valenta, Vaclav

    2014-03-01

    This paper summarizes the basis for the solution of heat removal by natural convection from both conventional nuclear reactors and reactors with fuel flowing coolant (such as reactors with molten fluoride salts MSR).The possibility of intensification of heat removal through gas lift is focused on. It might be used in an MSR (Molten Salt Reactor) for cleaning the salt mixture of degassed fission products and therefore eliminating problems with iodine pitting. Heat removal by natural convection and its intensification increases significantly the safety of nuclear reactors. Simultaneously the heat removal also solves problems with lifetime of pumps in the primary circuit of high-temperature reactors.

  5. 75 FR 58449 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Materials...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-09-24

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Materials, Metallurgy & Reactor Fuels The ACRS Subcommittee on Materials, Metallurgy & Reactor... would result in a major inconvenience. Dated: September 17, 2010. Antonio Dias, Chief, Reactor Safety...

  6. HWCTR CONTROL ROD AND SAFETY ROD DRIVE SYSTEMS

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Kale, S.H.

    1963-07-01

    The Heavy Water Components Test Reactor (HWCTR) is a pressurized, D/sub 2/O reactor designed for operation up to 70 Mw at 1500 psig and 3l5 deg C. It has 18 control rods and six safety rods, each driven by an electric motor through a rack and pinion gear train. Racks, pinions, and bearings are located inside individual pressure housings that are penetrated by means of floating ring labyrinth seals. The drives are mounted on the reactor vessel top head. Safety rods have electromagnetic clutches and fall into the reactor when scrammed. The reliability and performance of the rod drives aremore » very good. Seal leakage is well within design limits. Recent inspections of seals and control rod plants showed no evidence of crud buildup or stress corrosion cracking of type 17- 4PH'' stainless steel components. (auth)« less

  7. Compendium of Traffic Safety Research Projects: 1985-1995 and Beyond

    DOT National Transportation Integrated Search

    1996-03-01

    The National Highway Traffic Safety Administration's Office of Program : Development and Evaluation (OPDE) conducts research and evaluation projects : dealing with human attitudes, behaviors and failures (motor vehicle crashes). : OPDE's focus is on ...

  8. Reactor control rod timing system

    DOEpatents

    Wu, Peter T. K.

    1982-01-01

    A fluid driven jet-edge whistle timing system for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

  9. Engineered nanomaterials: toward effective safety management in research laboratories.

    PubMed

    Groso, Amela; Petri-Fink, Alke; Rothen-Rutishauser, Barbara; Hofmann, Heinrich; Meyer, Thierry

    2016-03-15

    It is still unknown which types of nanomaterials and associated doses represent an actual danger to humans and environment. Meanwhile, there is consensus on applying the precautionary principle to these novel materials until more information is available. To deal with the rapid evolution of research, including the fast turnover of collaborators, a user-friendly and easy-to-apply risk assessment tool offering adequate preventive and protective measures has to be provided. Based on new information concerning the hazards of engineered nanomaterials, we improved a previously developed risk assessment tool by following a simple scheme to gain in efficiency. In the first step, using a logical decision tree, one of the three hazard levels, from H1 to H3, is assigned to the nanomaterial. Using a combination of decision trees and matrices, the second step links the hazard with the emission and exposure potential to assign one of the three nanorisk levels (Nano 3 highest risk; Nano 1 lowest risk) to the activity. These operations are repeated at each process step, leading to the laboratory classification. The third step provides detailed preventive and protective measures for the determined level of nanorisk. We developed an adapted simple and intuitive method for nanomaterial risk management in research laboratories. It allows classifying the nanoactivities into three levels, additionally proposing concrete preventive and protective measures and associated actions. This method is a valuable tool for all the participants in nanomaterial safety. The users experience an essential learning opportunity and increase their safety awareness. Laboratory managers have a reliable tool to obtain an overview of the operations involving nanomaterials in their laboratories; this is essential, as they are responsible for the employee safety, but are sometimes unaware of the works performed. Bringing this risk to a three-band scale (like other types of risks such as biological, radiation

  10. The current state of the Russian reduced enrichment research reactors program

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Aden, V.G.; Kartashov, E.F.; Lukichev, V.A.

    1997-08-01

    During the last year after the 16-th International Conference on Reducing Fuel Enrichment in Research Reactors held in October, 1993 in Oarai, Japan, the conclusive stage of the Program on reducing fuel enrichment (to 20% in U-235) in research reactors was finally made up in Russia. The Program was started late in 70th and the first stage of the Program was completed by 1986 which allowed to reduce fuel enrichment from 80-90% to 36%. The completion of the Program current stage, which is counted for 5-6 years, will exclude the use of the fuel enriched by more than 20% frommore » RF to other countries such as: Poland, Czeck Republick, Hungary, Roumania, Bulgaria, Libya, Viet-Nam, North Korea, Egypt, Latvia, Ukraine, Uzbekistan and Kazakhstan. In 1994 the Program, approved by RF Minatom authorities, has received the status of an inter-branch program since it was admitted by the RF Ministry for Science and Technical Policy. The Head of RF Minatom central administrative division N.I.Ermakov was nominated as the Head of the Russian Program, V.G.Aden, RDIPE Deputy Director, was nominated as the scientific leader. The Program was submitted to the Commission for Scientific, Technical and Economical Cooperation between USA and Russia headed by Vice-President A. Gore and Prime Minister V. Chemomyrdin and was given support also.« less

  11. Small space reactor power systems for unmanned solar system exploration missions

    NASA Technical Reports Server (NTRS)

    Bloomfield, Harvey S.

    1987-01-01

    A preliminary feasibility study of the application of small nuclear reactor space power systems to the Mariner Mark II Cassini spacecraft/mission was conducted. The purpose of the study was to identify and assess the technology and performance issues associated with the reactor power system/spacecraft/mission integration. The Cassini mission was selected because study of the Saturn system was identified as a high priority outer planet exploration objective. Reactor power systems applied to this mission were evaluated for two different uses. First, a very small 1 kWe reactor power system was used as an RTG replacement for the nominal spacecraft mission science payload power requirements while still retaining the spacecraft's usual bipropellant chemical propulsion system. The second use of reactor power involved the additional replacement of the chemical propulsion system with a small reactor power system and an electric propulsion system. The study also provides an examination of potential applications for the additional power available for scientific data collection. The reactor power system characteristics utilized in the study were based on a parametric mass model that was developed specifically for these low power applications. The model was generated following a neutronic safety and operational feasibility assessment of six small reactor concepts solicited from U.S. industry. This assessment provided the validation of reactor safety for all mission phases and generatad the reactor mass and dimensional data needed for the system mass model.

  12. Beyond Decision Making for Outdoor Leaders: Expanding the Safety Behavior Research Agenda

    ERIC Educational Resources Information Center

    Jackson, Jeff S.

    2016-01-01

    The study of safety behaviour of designated outdoor leaders primarily revolves around their decision making and judgement. The last ten years, however, have seen relatively little peer-reviewed research regarding guide or instructor safety cognition and behaviour. The narrow decision making focus of modern work makes for a field of study…

  13. Bridging the divide between fire safety research and fighting fire safely: How do we convey research innovation to contribute more effectively to wildland firefighter safety?

    Treesearch

    Theodore Ted Adams; Bret W. Butler; Sara Brown; Vita Wright; Anne Black

    2017-01-01

    Creating a safe workplace for wildland firefighters has long been at the centre of discussion for researchers and practitioners. The goal of wildland fire safety research has been to protect operational firefighters, yet its contributions often fall short of potential because much is getting lost in the translation of peer-reviewed results to potential and intended...

  14. Assessment of Contributions to Patient Safety Knowledge by the Agency for Healthcare Research and Quality-Funded Patient Safety Projects

    PubMed Central

    Sorbero, Melony E S; Ricci, Karen A; Lovejoy, Susan; Haviland, Amelia M; Smith, Linda; Bradley, Lily A; Hiatt, Liisa; Farley, Donna O

    2009-01-01

    Objective To characterize the activities of projects funded in Agency for Healthcare Research and Quality (AHRQ)' patient safety portfolio and assess their aggregate potential to contribute to knowledge development. Data Sources Information abstracted from proposals for projects funded in AHRQ' patient safety portfolio, information on safety practices from the AHRQ Evidence Report on Patient Safety Practices, and products produced by the projects. Study Design This represented one part of the process evaluation conducted as part of a longitudinal evaluation based on the Context–Input–Process–Product model. Principal Findings The 234 projects funded through AHRQ' patient safety portfolio examined a wide variety of patient safety issues and extended their work beyond the hospital setting to less studied parts of the health care system. Many of the projects implemented and tested practices for which the patient safety evidence report identified a need for additional evidence. The funded projects also generated a substantial body of new patient safety knowledge through a growing number of journal articles and other products. Conclusions The projects funded in AHRQ' patient safety portfolio have the potential to make substantial contributions to the knowledge base on patient safety. The full value of this new knowledge remains to be confirmed through the synthesis of results. PMID:21456108

  15. Risk Informed Margins Management as part of Risk Informed Safety Margin Characterization

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Curtis Smith

    2014-06-01

    The ability to better characterize and quantify safety margin is important to improved decision making about Light Water Reactor (LWR) design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margin management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. In addition, as research and development in the LWR Sustainability (LWRS) Program and other collaborative efforts yield new data, sensors, and improved scientific understanding of physical processes that govern the aging and degradation of plant SSCs needs and opportunities to better optimize plantmore » safety and performance will become known. To support decision making related to economics, readability, and safety, the Risk Informed Safety Margin Characterization (RISMC) Pathway provides methods and tools that enable mitigation options known as risk informed margins management (RIMM) strategies.« less

  16. 75 FR 66168 - Seeks Qualified Candidates for the Advisory Committee on Reactor Safeguards

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-10-27

    ... NUCLEAR REGULATORY COMMISSION Seeks Qualified Candidates for the Advisory Committee on Reactor... Reactor Safeguards (ACRS). Submit r[eacute]sum[eacute]s to Ms. Brandi Hamilton, ACRS, Mail Stop T2E-26, U... of existing and proposed nuclear power plants and on the adequacy of proposed reactor safety...

  17. Reviewing methodologically disparate data: a practical guide for the patient safety research field.

    PubMed

    Brown, Katrina F; Long, Susannah J; Athanasiou, Thanos; Vincent, Charles A; Kroll, J Simon; Sevdalis, Nick

    2012-02-01

    This article addresses key questions frequently asked by researchers conducting systematic reviews in patient safety. This discipline is relatively young, and asks complex questions about complex aspects of health care delivery and experience, therefore its studies are typically methodologically heterogeneous, non-randomized and complex; but content rich and highly relevant to practice. Systematic reviews are increasingly necessary to drive forward practice and research in this area, but the data do not always lend themselves to 'standard' review methodologies. This accessible 'how-to' article demonstrates that data diversity need not preclude high-quality systematic reviews. It draws together information from published guidelines and experience within our multidisciplinary patient safety research group to provide entry-level advice for the clinician-researcher new to systematic reviewing, to non-biomedical research data or to both. It offers entry-level advice, illustrated with detailed practical examples, on defining a research question, creating a comprehensive search strategy, selecting articles for inclusion, assessing study quality, extracting data, synthesizing data and evaluating the impact of your review. The article concludes with a comment on the vital role of robust systematic reviews in the continuing advancement of the patient safety field. © 2010 Blackwell Publishing Ltd.

  18. Characteristics and Dose Levels for Spent Reactor Fuels

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Coates, Cameron W

    2007-01-01

    Current guidance considers highly radioactive special nuclear materials to be those materials that, unshielded, emit a radiation dose [rate] measured at 1 m which exceeds 100 rem/h. Smaller, less massive fuel assemblies from research reactors can present a challenge from the point of view of self protection because of their size (lower dose, easier to handle) and the desirability of higher enrichments; however, a follow-on study to cross-compare dose trends of research reactors and power reactors was deemed useful to confirm/verify these trends. This paper summarizes the characteristics and dose levels of spent reactor fuels for both research reactors andmore » power reactors and extends previous studies aimed at quantifying expected dose rates from research reactor fuels worldwide.« less

  19. Accident Analysis for the NIST Research Reactor Before and After Fuel Conversion

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Baek J.; Diamond D.; Cuadra, A.

    Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a modelmore » of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.« less

  20. FFTF Passive Safety Test Data for Benchmarks for New LMR Designs

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Wootan, David W.; Casella, Andrew M.

    Liquid Metal Reactors (LMRs) continue to be considered as an attractive concept for advanced reactor design. Software packages such as SASSYS are being used to im-prove new LMR designs and operating characteristics. Significant cost and safety im-provements can be realized in advanced liquid metal reactor designs by emphasizing inherent or passive safety through crediting the beneficial reactivity feedbacks associ-ated with core and structural movement. This passive safety approach was adopted for the Fast Flux Test Facility (FFTF), and an experimental program was conducted to characterize the structural reactivity feedback. The FFTF passive safety testing pro-gram was developed to examine howmore » specific design elements influenced dynamic re-activity feedback in response to a reactivity input and to demonstrate the scalability of reactivity feedback results to reactors of current interest. The U.S. Department of En-ergy, Office of Nuclear Energy Advanced Reactor Technology program is in the pro-cess of preserving, protecting, securing, and placing in electronic format information and data from the FFTF, including the core configurations and data collected during the passive safety tests. Benchmarks based on empirical data gathered during operation of the Fast Flux Test Facility (FFTF) as well as design documents and post-irradiation examination will aid in the validation of these software packages and the models and calculations they produce. Evaluation of these actual test data could provide insight to improve analytical methods which may be used to support future licensing applications for LMRs« less

  1. A Systematic Review of Community Health Workers' Role in Occupational Safety and Health Research.

    PubMed

    Swanberg, Jennifer E; Nichols, Helen M; Clouser, Jessica M; Check, Pietra; Edwards, Lori; Bush, Ashley M; Padilla, Yancy; Betz, Gail

    2018-03-03

    We systematically reviewed the literature to describe how community health workers (CHWs) are involved in occupational health and safety research and to identify areas for future research and research practice strategies. We searched five electronic databases from July 2015 through July 2016. Inclusion criteria were as follows: (1) study took place in the United States, (2) published as a full peer-review manuscript in English, (3) conducted occupational health and safety research, and (4) CHWs were involved in the research. The majority of 17 included studies took place in the agriculture industry (76%). CHWs were often involved in study implementation/design and research participant contact. Rationale for CHW involvement in research was due to local connections/acceptance, existing knowledge/skills, communication ability, and access to participants. Barriers to CHW involvement in research included competing demands on CHWs, recruitment and training difficulties, problems about research rigor and issues with proper data collection. Involving CHWs in occupational health and safety research has potential for improving inclusion of diverse, vulnerable and geographically isolated populations. Further research is needed to assess the challenges and opportunities of involving CHWs in this research and to develop evidence-based training strategies to teach CHWs to be lay-health researchers.

  2. Filtered epithermal quasi-monoenergetic neutron beams at research reactor facilities.

    PubMed

    Mansy, M S; Bashter, I I; El-Mesiry, M S; Habib, N; Adib, M

    2015-03-01

    Filtered neutron techniques were applied to produce quasi-monoenergetic neutron beams in the energy range of 1.5-133keV at research reactors. A simulation study was performed to characterize the filter components and transmitted beam lines. The filtered beams were characterized in terms of the optimal thickness of the main and additive components. The filtered neutron beams had high purity and intensity, with low contamination from the accompanying thermal emission, fast neutrons and γ-rays. A computer code named "QMNB" was developed in the "MATLAB" programming language to perform the required calculations. Copyright © 2014 Elsevier Ltd. All rights reserved.

  3. Interface design of VSOP'94 computer code for safety analysis

    NASA Astrophysics Data System (ADS)

    Natsir, Khairina; Yazid, Putranto Ilham; Andiwijayakusuma, D.; Wahanani, Nursinta Adi

    2014-09-01

    Today, most software applications, also in the nuclear field, come with a graphical user interface. VSOP'94 (Very Superior Old Program), was designed to simplify the process of performing reactor simulation. VSOP is a integrated code system to simulate the life history of a nuclear reactor that is devoted in education and research. One advantage of VSOP program is its ability to calculate the neutron spectrum estimation, fuel cycle, 2-D diffusion, resonance integral, estimation of reactors fuel costs, and integrated thermal hydraulics. VSOP also can be used to comparative studies and simulation of reactor safety. However, existing VSOP is a conventional program, which was developed using Fortran 65 and have several problems in using it, for example, it is only operated on Dec Alpha mainframe platforms and provide text-based output, difficult to use, especially in data preparation and interpretation of results. We develop a GUI-VSOP, which is an interface program to facilitate the preparation of data, run the VSOP code and read the results in a more user friendly way and useable on the Personal 'Computer (PC). Modifications include the development of interfaces on preprocessing, processing and postprocessing. GUI-based interface for preprocessing aims to provide a convenience way in preparing data. Processing interface is intended to provide convenience in configuring input files and libraries and do compiling VSOP code. Postprocessing interface designed to visualized the VSOP output in table and graphic forms. GUI-VSOP expected to be useful to simplify and speed up the process and analysis of safety aspects.

  4. A study of passive safety features by utilizing intra-subassembly-equipped self-actuated shutdown mechanism for future large fast breeder reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Uto, N.; Niwa, H.; Ieda, Y.

    1996-08-01

    Passive prevention of core disruptive accidents (CDAs) is desired in terms of enhancement of safety for future fast breeder reactors. In addition, mitigation of CDA`s consequences should be required because mitigation measures have a potential of applying to all accidents, while prevention measures are prepared for specific accident initiators. In this paper, the Intra-Subassembly-equipped Self-Actuated Shutdown System (IS-SASS) , which is considered effective on passive prevention and mitigation of CDAs, is described. The IS-SASS is introduced in a fuel subassembly and consists of absorber materials at the top of the active core and an inner duct through which molten fuelmore » can be excluded out of the core. The determination of the appropriate number of the IS-SASS units, their arrangement in the core and their suitable structure are found to be suited to prevention and mitigation of CDAs for liquid metal-cooled large fast breeder reactors.« less

  5. Fault tree applications within the safety program of Idaho Nuclear Corporation

    NASA Technical Reports Server (NTRS)

    Vesely, W. E.

    1971-01-01

    Computerized fault tree analyses are used to obtain both qualitative and quantitative information about the safety and reliability of an electrical control system that shuts the reactor down when certain safety criteria are exceeded, in the design of a nuclear plant protection system, and in an investigation of a backup emergency system for reactor shutdown. The fault tree yields the modes by which the system failure or accident will occur, the most critical failure or accident causing areas, detailed failure probabilities, and the response of safety or reliability to design modifications and maintenance schemes.

  6. Policies and practices pertaining to the selection, qualification requirements, and training programs for nuclear-reactor operating personnel at the Oak Ridge National Laboratory

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Culbert, W.H.

    1985-10-01

    This document describes the policies and practices of the Oak Ridge National Laboratory (ORNL) regarding the selection of and training requirements for reactor operating personnel at the Laboratory's nuclear-reactor facilities. The training programs, both for initial certification and for requalification, are described and provide the guidelines for ensuring that ORNL's research reactors are operated in a safe and reliable manner by qualified personnel. This document gives an overview of the reactor facilities and addresses the various qualifications, training, testing, and requalification requirements stipulated in DOE Order 5480.1A, Chapter VI (Safety of DOE-Owned Reactors); it is intended to be in compliancemore » with this DOE Order, as applicable to ORNL facilities. Included also are examples of the documentation maintained amenable for audit.« less

  7. MCNP-model for the OAEP Thai Research Reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Gallmeier, F.X.; Tang, J.S.; Primm, R.T. III

    An MCNP input was prepared for the Thai Research Reactor, making extensive use of the MCNP geometry`s lattice feature that allows a flexible and easy rearrangement of the core components and the adjustment of the control elements. The geometry was checked for overdefined or undefined zones by two-dimensional plots of cuts through the core configuration with the MCNP geometry plotting capabilities, and by a three-dimensional view of the core configuration with the SABRINA code. Cross sections were defined for a hypothetical core of 67 standard fuel elements and 38 low-enriched uranium fuel elements--all filled with fresh fuel. Three test calculationsmore » were performed with the MCNP4B-code to obtain the multiplication factor for the cases with control elements fully inserted, fully withdrawn, and at a working position.« less

  8. Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors

    NASA Astrophysics Data System (ADS)

    Scarlat, Raluca Olga

    This dissertation treats system design, modeling of transient system response, and characterization of individual phenomena and demonstrates a framework for integration of these three activities early in the design process of a complex engineered system. A system analysis framework for prioritization of experiments, modeling, and development of detailed design is proposed. Two fundamental topics in thermal-hydraulics are discussed, which illustrate the integration of modeling and experimentation with nuclear reactor design and safety analysis: thermal-hydraulic modeling of heat generating pebble bed cores, and scaled experiments for natural circulation heat removal with Boussinesq liquids. The case studies used in this dissertation are derived from the design and safety analysis of a pebble bed fluoride salt cooled high temperature nuclear reactor (PB-FHR), currently under development in the United States at the university and national laboratories level. In the context of the phenomena identification and ranking table (PIRT) methodology, new tools and approaches are proposed and demonstrated here, which are specifically relevant to technology in the early stages of development, and to analysis of passive safety features. A system decomposition approach is proposed. Definition of system functional requirements complements identification and compilation of the current knowledge base for the behavior of the system. Two new graphical tools are developed for ranking of phenomena importance: a phenomena ranking map, and a phenomena identification and ranking matrix (PIRM). The functional requirements established through this methodology were used for the design and optimization of the reactor core, and for the transient analysis and design of the passive natural circulation driven decay heat removal system for the PB-FHR. A numerical modeling approach for heat-generating porous media, with multi-dimensional fluid flow is presented. The application of this modeling

  9. Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Not Available

    1992-04-01

    This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

  10. Determinants of safety outcomes and performance: A systematic literature review of research in four high-risk industries.

    PubMed

    Cornelissen, Pieter A; Van Hoof, Joris J; De Jong, Menno D T

    2017-09-01

    In spite of increasing governmental and organizational efforts, organizations still struggle to improve the safety of their employees as evidenced by the yearly 2.3 million work-related deaths worldwide. Occupational safety research is scattered and inaccessible, especially for practitioners. Through systematically reviewing the safety literature, this study aims to provide a comprehensive overview of behavioral and circumstantial factors that endanger or support employee safety. A broad search on occupational safety literature using four online bibliographical databases yielded 27.527 articles. Through a systematic reviewing process 176 online articles were identified that met the inclusion criteria (e.g., original peer-reviewed research; conducted in selected high-risk industries; published between 1980-2016). Variables and the nature of their interrelationships (i.e., positive, negative, or nonsignificant) were extracted, and then grouped and classified through a process of bottom-up coding. The results indicate that safety outcomes and performance prevail as dependent research areas, dependent on variables related to management & colleagues, work(place) characteristics & circumstances, employee demographics, climate & culture, and external factors. Consensus was found for five variables related to safety outcomes and seven variables related to performance, while there is debate about 31 other relationships. Last, 21 variables related to safety outcomes and performance appear understudied. The majority of safety research has focused on addressing negative safety outcomes and performance through variables related to others within the organization, the work(place) itself, employee demographics, and-to a lesser extent-climate & culture and external factors. This systematic literature review provides both scientists and safety practitioners an overview of the (under)studied behavioral and circumstantial factors related to occupational safety behavior. Scientists

  11. SP-100 reactor with Brayton conversion for lunar surface applications

    NASA Technical Reports Server (NTRS)

    Mason, Lee S.; Rodriguez, Carlos D.; Mckissock, Barbara I.; Hanlon, James C.; Mansfield, Brian C.

    1992-01-01

    Examined here is the potential for integrating Brayton-cycle power conversion with the SP-100 reactor for lunar surface power system applications. Two designs were characterized and modeled. The first design integrates a 100-kWe SP-100 Brayton power system with a lunar lander. This system is intended to meet early lunar mission power needs while minimizing on-site installation requirements. Man-rated radiation protection is provided by an integral multilayer, cylindrical lithium hydride/tungsten (LiH/W) shield encircling the reactor vessel. Design emphasis is on ease of deployment, safety, and reliability, while utilizing relatively near-term technology. The second design combines Brayton conversion with the SP-100 reactor in a erectable 550-kWe powerplant concept intended to satisfy later-phase lunar base power requirements. This system capitalizes on experience gained from operating the initial 100-kWe module and incorporates some technology improvements. For this system, the reactor is emplaced in a lunar regolith excavation to provide man-rated shielding, and the Brayton engines and radiators are mounted on the lunar surface and extend radially from the central reactor. Design emphasis is on performance, safety, long life, and operational flexibility.

  12. Final report, PT IP-535-C: Test of smaller VSR`s in DR reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Vaughn, A.D.

    1963-04-17

    Because of rod-sticking problems at DR Reactor, a knuckle rod of B Reactor design was installed in vertical safety channel number 28. The substitute VSR, which has a smaller diameter than the original DR rod, has been tested for its operational feasibility including both drop time and reactivity effect. The reactivity effect of the rod was estimated by comparison of the reactivity transient caused by insertion of the specific B-type rod after scramming into the pile, with similar transients caused by normal vertical safety rod in an adjacent channel. This document lists the indicated relative control strength of the rodmore » as an empirical basis for future safety calculations. Results indicate that the B-type knuckel rod in DR reactor is about 80% as strong as a normal DR vertical safety rod if used in equivalent flux distribution and location; this makes it reasonable to assume that the local control strength of the B-type knuckel rod is 98 {mu}b.« less

  13. EDITORIAL: Safety aspects of fusion power plants

    NASA Astrophysics Data System (ADS)

    Kolbasov, B. N.

    2007-07-01

    This special issue of Nuclear Fusion contains 13 informative papers that were initially presented at the 8th IAEA Technical Meeting on Fusion Power Plant Safety held in Vienna, Austria, 10-13 July 2006. Following recommendation from the International Fusion Research Council, the IAEA organizes Technical Meetings on Fusion Safety with the aim to bring together experts to discuss the ongoing work, share new ideas and outline general guidance and recommendations on different issues related to safety and environmental (S&E) aspects of fusion research and power facilities. Previous meetings in this series were held in Vienna, Austria (1980), Ispra, Italy (1983), Culham, UK (1986), Jackson Hole, USA (1989), Toronto, Canada (1993), Naka, Japan (1996) and Cannes, France (2000). The recognized progress in fusion research and technology over the last quarter of a century has boosted the awareness of the potential of fusion to be a practically inexhaustible and clean source of energy. The decision to construct the International Thermonuclear Experimental Reactor (ITER) represents a landmark in the path to fusion power engineering. Ongoing activities to license ITER in France look for an adequate balance between technological and scientific deliverables and complying with safety requirements. Actually, this is the first instance of licensing a representative fusion machine, and it will very likely shape the way in which a more common basis for establishing safety standards and policies for licensing future fusion power plants will be developed. Now that ITER licensing activities are underway, it is becoming clear that the international fusion community should strengthen its efforts in the area of designing the next generations of fusion power plants—demonstrational and commercial. Therefore, the 8th IAEA Technical Meeting on Fusion Safety focused on the safety aspects of power facilities. Some ITER-related safety issues were reported and discussed owing to their potential

  14. Safety system augmentation at Russian nuclear power plants

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Scerbo, J.A.; Satpute, S.N.; Donkin, J.Y.

    1996-12-31

    This paper describes the design and procurement of a Class IE DC power supply system to upgrade plant safety at the Kola Nuclear Power Plant (NPP). Kola NPP is located above the Arctic circle at Polyarnie Zorie, Murmansk, Russia. Kola NPP consists of four units. Units 1 and 2 have VVER-440/230 type reactors: Units 3 and 4 have VVER-440/213 type reactors. The VVER-440 reactor design is similar to the pressurized water reactor design used in the US. This project provided redundant, Class 1E DC station batteries and DC switchboards for Kola NPP, Units 1 and 2. The new DC powermore » supply system was designed and procured in compliance with current nuclear design practices and requirements. Technical issues that needed to be addressed included reconciling the requirements in both US and Russian codes and satisfying the requirements of the Russian nuclear regulatory authority. Close interface with ATOMENERGOPROEKT (AEP), the Russian design organization, KOLA NPP plant personnel, and GOSATOMNADZOR (GAN), the Russian version of US Nuclear Regulatory Commission, was necessary to develop a design that would assure compliance with current Russian design requirements. Hence, this project was expected to serve as an example for plant upgrades at other similar VVER-440 nuclear plants. In addition to technical issues, the project needed to address language barriers and the logistics of shipping equipment to a remote section of the Former Soviet Union (FSU). This project was executed by Burns and Roe under the sponsorship of the US DOE as part of the International Safety Program (INSP). The INSP is a comprehensive effort, in cooperation with partners in other countries, to improve nuclear safety worldwide. A major element within the INSP is the improvement of the safety of Soviet-designed nuclear reactors.« less

  15. Perceptions about safety and risks in gender-based violence research: implications for the ethics review process.

    PubMed

    Sikweyiya, Yandisa; Jewkes, Rachel

    2011-10-01

    Does research on gender-based violence (GBV) pose greater than minimal risk to researchers and participants? This question needs to be understood particularly in light of hesitancy by Institutional Review Boards to approve research on GBV. The safety and risks of doing GBV studies and the implications for the ethical review process have not been a focus of much research. This qualitative study collected data through in-depth interviews with 12 experienced GBV researchers from various countries and a desk review. This paper explores researchers' interpretation of and meanings of the safety recommendations as provided in the WHO guidelines and whether there is empirical evidence on the presence of risks and safety concerns unique to GBV research. Informants raised a number of safety concerns about GBV research, yet in the interviews there were very few examples of problems having occurred, possibly because of the precautions applied. This paper argues that the notion that GBV studies carry greater than minimal risk when ethics precautions are followed is based on speculation, not evidence. It highlights the need for empirical evidence to support assertions of risk in research.

  16. 75 FR 57302 - Advisory Committee on Reactor Safeguards; Public Meeting

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-09-20

    ... Monday, October 14, 2009, (74 FR 52829-52830). Thursday, October 7, 2010, Conference Room T2-B1, Two....: Final Safety Evaluation Report Associated with the Economic Simplified Boiling Water Reactor (ESBWR..., Inc. regarding the final Safety Evaluation Report associated with the ESBWR design certification...

  17. A novel approach to enhance food safety: industry-academia-government partnership for applied research.

    PubMed

    Osterholm, Michael T; Ostrowsky, Julie; Farrar, Jeff A; Gravani, Robert B; Tauxe, Robert V; Buchanan, Robert L; Hedberg, Craig W

    2009-07-01

    An independent collaborative approach was developed for stimulating research on high-priority food safety issues. The Fresh Express Produce Safety Research Initiative was launched in 2007 with $2 million in unrestricted funds from industry and independent direction and oversight from a scientific advisory panel consisting of nationally recognized food safety experts from academia and government agencies. The program had two main objectives: (i) to fund rigorous, innovative, and multidisciplinary research addressing the safety of lettuce, spinach, and other leafy greens and (ii) to share research findings as widely and quickly as possible to support the development of advanced safeguards within the fresh-cut produce industry. Sixty-five proposals were submitted in response to a publicly announced request for proposals and were competitively evaluated. Nine research projects were funded to examine underlying factors involved in Escherichia coli O157:H7 contamination of lettuce, spinach, and other leafy greens and potential strategies for preventing the spread of foodborne pathogens. Results of the studies, published in the Journal of Food Protection, help to identify promising directions for future research into potential sources and entry points of contamination and specific factors associated with harvesting, processing, transporting, and storing produce that allow contaminants to persist and proliferate. The program provides a model for leveraging the strengths of industry, academia, and government to address high-priority issues quickly and directly through applied research. This model can be productively extended to other pathogens and other leafy and nonleafy produce.

  18. A review of pedestrian safety research in the United States and abroad

    DOT National Transportation Integrated Search

    2003-11-01

    The purpose of this report is to provide an overview of research studies on pedestrian safety in the United States; some foreign research also is included. Readers will find details of pedestrian crash characteristics, measures of pedestrian exposure...

  19. Research and proposal on selective catalytic reduction reactor optimization for industrial boiler.

    PubMed

    Yang, Yiming; Li, Jian; He, Hong

    2017-08-24

    The advanced computational fluid dynamics (CFD) software STAR-CCM+ was used to simulate a denitrification (De-NOx) project for a boiler in this paper, and the simulation result was verified based on a physical model. Two selective catalytic reduction (SCR) reactors were developed: reactor 1 was optimized and reactor 2 was developed based on reactor 1. Various indicators, including gas flow field, ammonia concentration distribution, temperature distribution, gas incident angle, and system pressure drop were analyzed. The analysis indicated that reactor 2 was of outstanding performance and could simplify developing greatly. Ammonia injection grid (AIG), the core component of the reactor, was studied; three AIGs were developed and their performances were compared and analyzed. The result indicated that AIG 3 was of the best performance. The technical indicators were proposed for SCR reactor based on the study. Flow filed distribution, gas incident angle, and temperature distribution are subjected to SCR reactor shape to a great extent, and reactor 2 proposed in this paper was of outstanding performance; ammonia concentration distribution is subjected to ammonia injection grid (AIG) shape, and AIG 3 could meet the technical indicator of ammonia concentration without mounting ammonia mixer. The developments above on the reactor and the AIG are both of great application value and social efficiency.

  20. 2003 highway-rail grade crossing safety research needs workshop. Volume 2 : appendices

    DOT National Transportation Integrated Search

    2009-01-01

    The purposes of the workshop were to provide up-to-date information and research reports from selected organizations, analyze a number of safety research topics by a selected group of delegates from all areas of technology and government organization...

  1. Healthcare quality and safety: a review of policy, practice and research.

    PubMed

    Waring, Justin; Allen, Davina; Braithwaite, Jeffrey; Sandall, Jane

    2016-02-01

    Over the last two decades healthcare quality and safety have risen to the fore of health policy and research. This has largely been informed by theoretical and empirical ideas found in the fields of ergonomics and human factors. These have enabled significant advances in our understanding and management of quality and safety. However, a parallel and at time neglected sociological literature on clinical quality and safety is presented as offering additional, complementary, and at times critical insights on the problems of quality and safety. This review explores the development and contributions of both the mainstream and more sociological approaches to safety. It shows that where mainstream approaches often focus on the influence of human and local environment factors in shaping quality, a sociological perspective can deepen knowledge of the wider social, cultural and political factors that contextualise the clinical micro-system. It suggests these different perspectives can easily complement one another, offering a more developed and layered understanding of quality and safety. It also suggests that the sociological literature can bring to light important questions about the limits of the more mainstream approaches and ask critical questions about the role of social inequality, power and control in the framing of quality and safety. © 2015 Foundation for the Sociology of Health & Illness.

  2. Modular fabrication and characterization of complex silicon carbide composite structures Advanced Reactor Technologies (ART) Research Final Report (Feb 2015 – May 2017)

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Khalifa, Hesham

    Advanced ceramic materials exhibit properties that enable safety and fuel cycle efficiency improvements in advanced nuclear reactors. In order to fully exploit these desirable properties, new processing techniques are required to produce the complex geometries inherent to nuclear fuel assemblies and support structures. Through this project, the state of complex SiC-SiC composite fabrication for nuclear components has advanced significantly. New methods to produce complex SiC-SiC composite structures have been demonstrated in the form factors needed for in-core structural components in advanced high temperature nuclear reactors. Advanced characterization techniques have been employed to demonstrate that these complex SiC-SiC composite structures providemore » the strength, toughness and hermeticity required for service in harsh reactor conditions. The complex structures produced in this project represent a significant step forward in leveraging the excellent high temperature strength, resistance to neutron induced damage, and low neutron cross section of silicon carbide in nuclear applications.« less

  3. Statewide traffic safety study phase I : review of current traffic safety research, practice, analytical procedures and databases.

    DOT National Transportation Integrated Search

    2005-05-01

    This report synthesized the research findings of Phase I of the Statewide Traffic Safety Study of Louisiana, sponsored by the Louisiana Department of Transportation and Development. The objective of Phase I was to provide a comprehensive review of th...

  4. Analysis of reactor trips originating in balance of plant systems

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Stetson, F.T.; Gallagher, D.W.; Le, P.T.

    1990-09-01

    This report documents the results of an analysis of balance-of-plant (BOP) related reactor trips at commercial US nuclear power plants of a 5-year period, from January 1, 1984, through December 31, 1988. The study was performed for the Plant Systems Branch, Office of Nuclear Reactor Regulation, US Nuclear Regulatory Commission. The objectives of the study were: to improve the level of understanding of BOP-related challenges to safety systems by identifying and categorizing such events; to prepare a computerized data base of BOP-related reactor trip events and use the data base to identify trends and patterns in the population of thesemore » events; to investigate the risk implications of BOP events that challenge safety systems; and to provide recommendations on how to address BOP-related concerns in regulatory context. 18 refs., 2 figs., 27 tabs.« less

  5. Food Safety at Farmers' Markets: A Knowledge Synthesis of Published Research.

    PubMed

    Young, Ian; Thaivalappil, Abhinand; Reimer, Danielle; Greig, Judy

    2017-12-01

    Farmers' markets are increasingly popular venues in North America for the sale of fresh produce and other foods. However, the nature of their operation can present possible food safety issues, challenges, and risks to consumers. A knowledge synthesis was conducted to identify, characterize, and summarize published research on the microbial food safety issues and implications associated with farmers' markets. A scoping review was conducted using the following steps: comprehensive search strategy, relevance screening of abstracts, and characterization of relevant articles. Two subsets of data were prioritized for more detailed systematic review (data extraction and risk-of-bias assessment) and meta-analysis: (i) studies comparing the microbial safety of foods from farmers' markets versus other sources and (ii) studies evaluating the use of food safety practices at farmers' markets. Overall, 83 relevant studies were identified. The majority of studies were published as journal articles (64%), used a cross-sectional design (81%), and were conducted in the United States (78%). Most studies (39%; n = 32) investigated stakeholder, mostly consumer (n = 22), attitudes toward food safety at farmers' markets. Limited but heterogeneous evidence indicated a higher prevalence of Campylobacter and Salmonella in chicken meat from farmers' markets versus other retail sources, but there was no difference in the microbial contamination of fresh produce. Studies evaluating the use of food safety practices at farmers' markets identified some gaps; for example, the average prevalence of vendor hand washing was 4% (95% confidence interval: 0 to 11%; I 2 = 27%; n = 5 studies). Twelve foodborne outbreaks and case reports were identified, resulting in a total of 411 illnesses, 38 hospitalizations, and two deaths from 1994 to 2016. Only five intervention studies were identified. Key knowledge gaps and areas warranting future research, training, and education are highlighted and discussed.

  6. 75 FR 62892 - Massachusetts Institute of Technology Research Reactor Environmental Assessment and Finding of No...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-10-13

    ... NUCLEAR REGULATORY COMMISSION [Docket No. 50-020; NRC-2010-0313] Massachusetts Institute of Technology Research Reactor Environmental Assessment and Finding of No Significant Impact Correction In notice document 2010-24809 beginning on page 61220 in the issue of Monday, October 4, 2010, make the...

  7. Research and development of a safety standard for workstation tables in the United States

    DOT National Transportation Integrated Search

    2013-02-21

    The US safety standard for workstation tables is presented to an international audience, : such that rail operators and equipment manufacturers may better understand the research : behind the requirements, the process through which the safety standar...

  8. Laser safety research and modeling for high-energy laser systems

    NASA Astrophysics Data System (ADS)

    Smith, Peter A.; Montes de Oca, Cecilia I.; Kennedy, Paul K.; Keppler, Kenneth S.

    2002-06-01

    The Department of Defense has an increasing number of high-energy laser weapons programs with the potential to mature in the not too distant future. However, as laser systems with increasingly higher energies are developed, the difficulty of the laser safety problem increases proportionally, and presents unique safety challenges. The hazard distance for the direct beam can be in the order of thousands of miles, and radiation reflected from the target may also be hazardous over long distances. This paper details the Air Force Research Laboratory/Optical Radiation Branch (AFRL/HEDO) High-Energy Laser (HEL) safety program, which has been developed to support DOD HEL programs by providing critical capability and knowledge with respect to laser safety. The overall aim of the program is to develop and demonstrate technologies that permit safe testing, deployment and use of high-energy laser weapons. The program spans the range of applicable technologies, including evaluation of the biological effects of high-energy laser systems, development and validation of laser hazard assessment tools, and development of appropriate eye protection for those at risk.

  9. Reactor design and integration into a nuclear electric spacecraft

    NASA Technical Reports Server (NTRS)

    Phillips, W. M.; Koenig, D. R.

    1978-01-01

    One of the well-defined applications for nuclear power in space is nuclear electric propulsion (NEP). Mission studies have identified the optimum power level (400 kWe). A single Shuttle launch requirement and science-package integration have added additional constraints to the design. A reactor design which will meet these constraints has been studied. The reactor employs 90 fuel elements, each heat pipe cooled. Reactor control is obtained with BeO/B4C drums in a BeO reflector. The balance of the spacecraft is shielded from the reactor with LiH. Power conditioning and reactor control drum drives are located behind the LiH with the power conditioning. Launch safety, mechanical design and integration with the power conversion subsystem are discussed.

  10. The structure and emerging trends of construction safety management research: a bibliometric review.

    PubMed

    Liang, Huakang; Zhang, Shoujian; Su, Yikun

    2018-03-29

    Recently, construction safety management (CSM) practices and systems have become important topics for stakeholders to take care of human resources. However, few studies have attempted to map the global research on CSM. A comprehensive bibliometric review was conducted in this study based on multiple methods. In total, 1172 CSM-related papers from the Web of Science Core Collection database were examined. The analyses focused on publication year, country-institute, publication source, author and research topics. The results indicated that the USA, China, Australia and the UK took leading positions in CSM research. Two branches of journals were identified, namely the branch of engineering science and that of safety science and social science. Additionally, seven themes together with 28 specific topics were detected to allow researchers to track the main structure and temporal evolution of CSM research. Finally, the main research trends and potential research directions were discussed to guide the future research.

  11. Development of a New Transportation/Storage Cask System for Use by the DOE Russian Research Reactor Fuel Return Program

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Michael J. Tyacke; Frantisek Svitak; Jiri Rychecky

    2007-10-01

    The United States, the Russian Federation, and the International Atomic Energy Agency (IAEA) have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program. The purpose of this program is to return Soviet or Russian-supplied high-enriched uranium (HEU) fuel, currently stored at Russian-designed research reactors throughout the world, to Russia. To accommodate transport of the HEU spent nuclear fuel (SNF), a new large-capacity transport/storage cask system was specially designed for handling and operations under the unique conditions at these research reactor facilities. This new cask system is named the ŠKODA VPVR/M cask. The design, licensing,more » testing, and delivery of this new cask system result from a significant international cooperative effort by several countries and involved numerous private and governmental organizations. This paper contains the following sections: 1) Introduction; 2) VPVR/M Cask Description; 3) Ancillary Equipment, 4) Cask Licensing; 5) Cask Demonstration and Operations; 6) IAEA Procurement, Quality Assurance Inspections, Fabrication, and Delivery; and, 7) Conclusions.« less

  12. Low exchange element for nuclear reactor

    DOEpatents

    Brogli, Rudolf H.; Shamasunder, Bangalore I.; Seth, Shivaji S.

    1985-01-01

    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. Nuclear reactor control apparatus

    DOEpatents

    Sridhar, Bettadapur N.

    1983-10-25

    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.

  14. Evaluation of Nuclear Facility Decommissioning Projects program: a reference research reactor. Project summary report

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Baumann, B.L.; Miller, R.L.

    1983-10-01

    This document presents, in summary form, generic conceptual information relevant to the decommissioning of a reference research reactor (RRR). All of the data presented were extracted from NUREG/CR-1756 and arranged in a form that will provide a basis for future comparison studies for the Evaluation of Nuclear Facility Decommissioning Projects (ENFDP) program.

  15. Reactor control rod timing system. [LMFBR

    DOEpatents

    Wu, P.T.K.

    1980-03-18

    A fluid driven jet-edge whistle timing system is described for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

  16. Development of the cascade inertial-confinement-fusion reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Pitts, J.H.

    Caqscade, originally conceived as a football-shaped, steel-walled reactor containing a Li/sub 2/O granule blanket, is now envisaged as a double-cone-shaped reactor containing a two-layered (three-zone) flowing blanket of BeO and LiAlO/sub 2/ granules. Average blanket exit temperature is 1670 K and gross plant efficiency (net thermal conversion efficiency) using a Brayton cycle is 55%. The reactor has a low-activation SiC-tiled wall. It rotates at 50 rpm, and the granules are transported to the top of the heat exchanger using their peripheral speed; no conveyors or lifts are required. The granules return to the reactor by gravity. After considerable analysis andmore » experimentation, we continue to regard Cascade as a promising reactor concept with the advantages of safety, efficiency, and low activation.« less

  17. Development of the cascade inertial-confinement-fusion reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Pitts, J.H.

    Cascade, originally conceived as a football-shaped, steel-walled reactor containing a Li/sub 2/O granule blanket, is now envisaged as a double-cone-shaped reactor containing a two-layered (three-zone) flowing blanket of BeO and LiAlO/sub 2/ granules. Average blanket exit temperature is 1670/sup 0/K and gross plant efficiency (net thermal conversion efficiency) using a Brayton cycle is 55%. The reactor has a low-activation SiC-tiled wall. It rotates at 50 rpm, and the granules are transported to the top of the heat exchanger using their peripheral speed; no conveyors or lifts are required. The granules return to the reactor by gravity. After considerable analysis andmore » experimentation, we continue to regard Cascade as a promising reactor concept with the advantages of safety, efficiency, and low activation.« less

  18. The Use Of High Speed Photography In Reactor Safety Studies At The Atomic Energy Establishment, Winfrith

    NASA Astrophysics Data System (ADS)

    Maddison, R. J.

    1985-02-01

    The investigation of certain areas of nuclear reactor safety involves the study of high speed phenomena with timescales ranging from microseconds to a few hundreds of milliseconds. Examples which have been extensively studied at Winfrith are firstly, the thermal interaction of molten fuel and reactor coolant which can generate high pressures on the 100 msec timescale, and which involves phenomena such as vapour film collapse which takes place on the microsecond timescale. Secondly, there is the response of reactor structures to such pressures, and finally there is the response of structural materials such as metals and concrete to the impulsive loading arising from the impact of heavy, high velocity missiles. A wide range of experimental techniques is used in these studies, many of which have been developed specially for this type of work which ranges from small laboratory scale to large field scale experiments. There are two important features which characterise many of these experiments:- i) a long period of meticulous preparation of very heavily instrumented, short duration experiments and; ii) the destructive nature of the experiments. Various forms of High Speed photography are included in the inventory of experimental techniques. These include the use of single and double exposure, short duration, spark photography; the use of an Image Convertor Camera (IMACON 790); and a number of rotating prism cine cameras. High Speed Photography is used both in a primary experimental role in the studies, and in a supportive role for other instrumentation. Because of the sometimes violent nature of these experiments, cameras are often heavily protected and operated remotely; lighting systems are sometimes destroyed. This has led to the development of unconventional techniques for camera operation and subject lighting. This paper will describe some of the experiments and the way in which High Speed Photography has been applied as an essential experimental tool. It will be

  19. Nuclear safety policy working group recommendations on nuclear propulsion safety for the space exploration initiative

    NASA Technical Reports Server (NTRS)

    Marshall, Albert C.; Lee, James H.; Mcculloch, William H.; Sawyer, J. Charles, Jr.; Bari, Robert A.; Cullingford, Hatice S.; Hardy, Alva C.; Niederauer, George F.; Remp, Kerry; Rice, John W.

    1993-01-01

    An interagency Nuclear Safety Working Group (NSPWG) was chartered to recommend nuclear safety policy, requirements, and guidelines for the Space Exploration Initiative (SEI) nuclear propulsion program. These recommendations, which are contained in this report, should facilitate the implementation of mission planning and conceptual design studies. The NSPWG has recommended a top-level policy to provide the guiding principles for the development and implementation of the SEI nuclear propulsion safety program. In addition, the NSPWG has reviewed safety issues for nuclear propulsion and recommended top-level safety requirements and guidelines to address these issues. These recommendations should be useful for the development of the program's top-level requirements for safety functions (referred to as Safety Functional Requirements). The safety requirements and guidelines address the following topics: reactor start-up, inadvertent criticality, radiological release and exposure, disposal, entry, safeguards, risk/reliability, operational safety, ground testing, and other considerations.

  20. Decommissioning of the Dragon High Temperature Reactor (HTR) Located at the Former United Kingdom Atomic Energy Authority (UKAEA) Research Site at Winfrith - 13180

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Smith, Anthony A.

    2013-07-01

    The Dragon Reactor was constructed at the United Kingdom Atomic Energy Research Establishment at Winfrith in Dorset through the late 1950's and into the early 1960's. It was a High Temperature Gas Cooled Reactor (HTR) with helium gas coolant and graphite moderation. It operated as a fuel testing and demonstration reactor at up to 20 MW (Thermal) from 1964 until 1975, when international funding for this project was terminated. The fuel was removed from the core in 1976 and the reactor was put into Safestore. To meet the UK's Nuclear Decommissioning Authority (NDA) objective to 'drive hazard reduction' [1] itmore » is necessary to decommission and remediate all the Research Sites Restoration Ltd (RSRL) facilities. This includes the Dragon Reactor where the activated core, pressure vessel and control rods and the contaminated primary circuit (including a {sup 90}Sr source) still remain. It is essential to remove these hazards at the appropriate time and return the area occupied by the reactor to a safe condition. (author)« less

  1. Status of DOE efforts to renew acceptance of foreign research reactor spent nuclear fuel

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Head, C.R.

    1997-08-01

    This presentation summarizes the efforts being made by the Department of Energy to renew acceptance of spent nuclear fuel shipments from foreign research reactors. The author reviews the actions undertaken in this process in a fairly chronological manner, through the present time, as well as the development of an environmental impact statement to support the proposed actions.

  2. NEUTRONIC REACTOR WITH ACCESSIBLE THIMBLE AND EMERGENCY COOLING FEATURES

    DOEpatents

    McCorkle, W.H.

    1960-02-23

    BS>A safety system for a water-moderated reactor is described. The invention comprises a reservoir system for spraying the fuel elements within a fuel assembly with coolant and keeping them in a continuous bath even if the coolant moderator is lost from the reactor vessel. A reservoir gravity feeds one or more nozzels positioned within each fuel assembly which continually forces water past the fuel elements.

  3. Core Dynamics Analysis for Reactivity Insertion and Loss of Coolant Flow Tests Using the High Temperature Engineering Test Reactor

    NASA Astrophysics Data System (ADS)

    Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

    Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-temperature Gas-cooled Reactors (HTGRs). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named-ACCORD-, was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We have modified this code to use a model with four parallel channels and twenty temperature coefficients. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results. Moreover, the effect of the model is formulated quantitatively with our proposed equation. Finally, the pre-analytical result of the loss of coolant flow test by tripping all gas circulators is also discussed.

  4. Interim status report on lead-cooled fast reactor (LFR) research and development.

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Tzanos, C. P.; Sienicki, J. J.; Moisseytsev, A.

    2008-03-31

    This report discusses the status of Lead-Cooled Fast Reactor (LFR) research and development carried out during the first half of FY 2008 under the U.S. Department of Energy Generation IV Nuclear Energy Systems Initiative. Lead-Cooled Fast Reactor research and development has recently been transferred from Generation IV to the Reactor Campaign of the Global Nuclear Energy Partnership (GNEP). Another status report shall be issued at the end of FY 2008 covering all of the LFR activities carried out in FY 2008 for both Generation IV and GNEP. The focus of research and development in FY 2008 is an initial investigationmore » of a concept for a LFR Advanced Recycling Reactor (ARR) Technology Pilot Plant (TPP)/demonstration test reactor (demo) incorporating features and operating conditions of the European Lead-cooled SYstem (ELSY) {approx} 600 MWe lead (Pb)-cooled LFR preconceptual design for the transmutation of waste and central station power generation, and which would enable irradiation testing of advanced fuels and structural materials. Initial scoping core concept development analyses have been carried out for a 100 MWt core composed of sixteen open-lattice 20 by 20 fuel assemblies largely similar to those of the ELSY preconceptual fuel assembly design incorporating fuel pins with mixed oxide (MOX) fuel, central control rods in each fuel assembly, and cooled with Pb coolant. For a cycle length of three years, the core is calculated to have a conversion ratio of 0.79, an average discharge burnup of 108 MWd/kg of heavy metal, and a burnup reactivity swing of about 13 dollars. With a control rod in each fuel assembly, the reactivity worth of an individual rod would need to be significantly greater than one dollar which is undesirable for postulated rod withdrawal reactivity insertion events. A peak neutron fast flux of 2.0 x 10{sup 15} (n/cm{sup 2}-s) is calculated. For comparison, the 400 MWt Fast Flux Test Facility (FFTF) achieved a peak neutron fast flux of 7.2 x

  5. Exploration of Heterogeneity in Distributed Research Network Drug Safety Analyses

    ERIC Educational Resources Information Center

    Hansen, Richard A.; Zeng, Peng; Ryan, Patrick; Gao, Juan; Sonawane, Kalyani; Teeter, Benjamin; Westrich, Kimberly; Dubois, Robert W.

    2014-01-01

    Distributed data networks representing large diverse populations are an expanding focus of drug safety research. However, interpreting results is difficult when treatment effect estimates vary across datasets (i.e., heterogeneity). In a previous study, risk estimates were generated for selected drugs and potential adverse outcomes. Analyses were…

  6. Assessment and Application of the ROSE Code for Reactor Outage Thermal-Hydraulic and Safety Analysis

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Liang, Thomas K.S.; Ko, F.-K.; Dai, L.-C

    The currently available tools, such as RELAP5, RETRAN, and others, cannot easily and correctly perform the task of analyzing the system behavior during plant outages. Therefore, a medium-sized program aiming at reactor outage simulation and evaluation, such as midloop operation (MLO) with loss of residual heat removal (RHR), has been developed. Important thermal-hydraulic processes involved during MLO with loss of RHR can be properly simulated by the newly developed reactor outage simulation and evaluation (ROSE) code. The two-region approach with a modified two-fluid model has been adopted to be the theoretical basis of the ROSE code.To verify the analytical modelmore » in the first step, posttest calculations against the integral midloop experiments with loss of RHR have been performed. The excellent simulation capacity of the ROSE code against the Institute of Nuclear Energy Research Integral System Test Facility test data is demonstrated. To further mature the ROSE code in simulating a full-sized pressurized water reactor, assessment against the WGOTHIC code and the Maanshan momentary-loss-of-RHR event has been undertaken. The successfully assessed ROSE code is then applied to evaluate the abnormal operation procedure (AOP) with loss of RHR during MLO (AOP 537.4) for the Maanshan plant. The ROSE code also has been successfully transplanted into the Maanshan training simulator to support operator training. How the simulator was upgraded by the ROSE code for MLO will be presented in the future.« less

  7. Research Lasers and Air Traffic Safety: Issues, Concerns and Responsibilities of the Research Community

    NASA Technical Reports Server (NTRS)

    Nessler, Phillip J., Jr.

    1998-01-01

    The subject of outdoor use of lasers relative to air traffic has become a diverse and dynamic topic. During the past several decades, the use of lasers in outdoor research activities have increased significantly. Increases in the outdoor use of lasers and increases in air traffic densities have changed the levels of risk involved. To date there have been no documented incidents of air traffic interference from research lasers; however, incidents involving display lasers have shown a marked increase. As a result of the national response to these incidents, new concerns over lasers have arisen. Through the efforts of the SAE G-10T Laser Safety Hazards Subcommittee and the ANSI Z136.6 development committee, potential detrimental effects to air traffic beyond the traditional eye damage concerns have been identified. An increased emphasis from the Federal Aviation Administration (FAA), the Center for Devices and Radiological Hazards (CDRH), and the National Transportation Safety Board (NTSB) along with increased concern by the public have resulted in focused scrutiny of potential hazards presented by lasers. The research community needs to rethink the traditional methods of risk evaluation and application of protective measures. The best current approach to assure adequate protection of air traffic is the application of viable hazard and risk analysis and the use of validated protective measures. Standards making efforts and regulatory development must be supported by the research community to assure that reasonable measures are developed. Without input, standards and regulations can be developed that are not compatible with the needs of the research community. Finally, support is needed for the continued development and validation of protective measures.

  8. The Feasibility of Establishing Highway Safety Manpower Development and Research Centers at University-Level Institutions. Final Report, Volume I: Study Report.

    ERIC Educational Resources Information Center

    Chorness, Maury H.; And Others

    To examine the feasibility of establishing Highway Safety Manpower Development and Research (HSMDR) Centers at university-level institutions which would produce three types of manpower--safety specialists, safety professionals, and research manpower, previous National Highway Safety Bureau research studies and approximately 50 federally funded…

  9. 75 FR 10840 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-03-09

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling Water Reactor (ABWR); Notice of Meeting The ACRS Subcommittee on ABWR will... would result in major inconvenience. Dated: March 3, 2010. Antonio F. Dias, Chief, Reactor Safety Branch...

  10. Reactor power system deployment and startup

    NASA Technical Reports Server (NTRS)

    Wetch, J. R.; Nelin, C. J.; Britt, E. J.; Klein, G.

    1985-01-01

    This paper addresses issues that should receive further examination in the near-term as concept selection for development of a U.S. space reactor power system is approached. The issues include: the economics, practicality and system reliability associated with transfer of nuclear spacecraft from low earth shuttle orbits to operational orbits, via chemical propulsion versus nuclear electric propulsion; possible astronaut supervised reactor and nuclear electric propulsion startup in low altitude Shuttle orbit; potential deployment methods for nuclear powered spacecraft from Shuttle; the general public safety of low altitude startup and nuclear safe and disposal orbits; the question of preferred reactor power level; and the question of frozen versus molten alkali metal coolant during launch and deployment. These issues must be considered now because they impact the SP-100 concept selection, power level selection, weight and size limits, use of deployable radiators, reliability requirements, and economics, as well as the degree of need for and the urgency of developing space reactor power systems.

  11. Tritium Mitigation/Control for Advanced Reactor System

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Sun, Xiaodong; Christensen, Richard; Saving, John P.

    A tritium removal facility, which is similar to the design used for tritium recovery in fusion reactors, is proposed in this study for fluoride-salt-cooled high-temperature reactors (FHRs) to result in a two-loop FHR design with the elimination of an intermediate loop. Using this approach, an economic benefit can potentially be obtained by removing the intermediate loop, while the safety concern of tritium release can be mitigated. In addition, an intermediate heat exchanger (IHX) that can yield a similar tritium permeation rate to the production rate of 1.9 Ci/day in a 1,000 MWe PWR needs to be designed to prevent themore » residual tritium that is not captured in the tritium removal system from escaping into the power cycle and ultimately the environment. The main focus of this study is to aid the mitigation of tritium permeation issue from the FHR primary side to significantly reduce the concentration of tritium in the secondary side and the process heat application side (if applicable). The goal of the research is to propose a baseline FHR system without the intermediate loop. The specific objectives to accomplish the goals are: To estimate tritium permeation behavior in FHRs; To design a tritium removal system for FHRs; To meet the same tritium permeation level in FHRs as the tritium production rate of 1.9 Ci/day in 1,000 MWe PWRs; To demonstrate economic benefits of the proposed FHR system via comparing with the three-loop FHR system. The objectives were accomplished by designing tritium removal facilities, developing a tritium analysis code, and conducting an economic analysis. In the fusion reactor community, tritium extraction has been widely investigated and researched. Borrowing the experiences from the fusion reactor community, a tritium control and mitigation system was proposed. Based on mass transport theories, a tritium analysis code was developed, and the tritium behaviors were analyzed using the developed code. Tritium removal facilities were

  12. Non-animal approaches for consumer safety risk assessments: Unilever's scientific research programme.

    PubMed

    Carmichael, Paul; Davies, Michael; Dent, Matt; Fentem, Julia; Fletcher, Samantha; Gilmour, Nicola; MacKay, Cameron; Maxwell, Gavin; Merolla, Leona; Pease, Camilla; Reynolds, Fiona; Westmoreland, Carl

    2009-12-01

    Non-animal based approaches to risk assessment are now routinely used for assuring consumer safety for some endpoints (such as skin irritation) following considerable investment in developing and applying new methods over the past 20 years. Unilever's research programme into non-animal approaches for safety assessment is currently focused on the application of new technologies to risk assessments in the areas of skin allergy, cancer and general toxicity (including inhalation toxicity). In all of these areas, a long-term investment is essential to increase the scientific understanding of the underlying biological and chemical processes that we believe will ultimately form a sound basis for novel risk assessment approaches. Our research programme in these priority areas consists of in-house research as well as Unilever-sponsored academic research, involvement with EU-funded projects (e.g. Sens-it-iv, carcinoGENOMICS), participation in cross-industry collaborative research (e.g. COLIPA, EPAA) and ongoing involvement with other scientific initiatives on non-animal approaches to risk assessment (e.g. UK NC3Rs, US 'Human Toxicology Project' consortium). 2009 FRAME.

  13. Radio-toxicity of spent fuel of the advanced heavy water reactor.

    PubMed

    Anand, S; Singh, K D S; Sharma, V K

    2010-01-01

    The Advanced Heavy Water Reactor (AHWR) is a new power reactor concept being developed at Bhabha Atomic Research Centre, Mumbai. The reactor retains many desirable features of the existing Pressurised Heavy Water Reactor (PHWR), while incorporating new, advanced safety features. The reactor aims to utilise the vast thorium resources available in India. The reactor core will use plutonium as the make-up fuel, while breeding (233)U in situ. On account of this unique combination of fuel materials, the operational characteristics of the fuel as determined by its radioactivity, decay heat and radio-toxicity are being viewed with great interest. Radio-toxicity of the spent fuel is a measure of potential radiological hazard to the members of the public and also important from the ecological point of view. The radio-toxicity of the AHWR fuel is extremely high to start with, being approximately 10(4) times that of the fresh natural U fuel used in a PHWR, and continues to remain relatively high during operation and subsequent cooling. A unique feature of this fuel is the peak observed in its radio-toxicity at approximately 10(5) y of decay cooling. The delayed increase in fuel toxicity has been traced primarily to a build-up of (229)Th, (230)Th and (226)Ra. This phenomenon has been observed earlier for thorium-based fuels and is confirmed for the AHWR fuel. This paper presents radio-toxicity data for AHWR spent fuel up to a period of 10(6) y and the results are compared with the radio-toxicity of PHWR.

  14. A Gas-Cooled-Reactor Closed-Brayton-Cycle Demonstration with Nuclear Heating

    NASA Astrophysics Data System (ADS)

    Lipinski, Ronald J.; Wright, Steven A.; Dorsey, Daniel J.; Peters, Curtis D.; Brown, Nicholas; Williamson, Joshua; Jablonski, Jennifer

    2005-02-01

    A gas-cooled reactor may be coupled directly to turbomachinery to form a closed-Brayton-cycle (CBC) system in which the CBC working fluid serves as the reactor coolant. Such a system has the potential to be a very simple and robust space-reactor power system. Gas-cooled reactors have been built and operated in the past, but very few have been coupled directly to the turbomachinery in this fashion. In this paper we describe the option for testing such a system with a small reactor and turbomachinery at Sandia National Laboratories. Sandia currently operates the Annular Core Research Reactor (ACRR) at steady-state powers up to 4 MW and has an adjacent facility with heavy shielding in which another reactor recently operated. Sandia also has a closed-Brayton-Cycle test bed with a converted commercial turbomachinery unit that is rated for up to 30 kWe of power. It is proposed to construct a small experimental gas-cooled reactor core and attach this via ducting to the CBC turbomachinery for cooling and electricity production. Calculations suggest that such a unit could produce about 20 kWe, which would be a good power level for initial surface power units on the Moon or Mars. The intent of this experiment is to demonstrate the stable start-up and operation of such a system. Of particular interest is the effect of a negative temperature power coefficient as the initially cold Brayton gas passes through the core during startup or power changes. Sandia's dynamic model for such a system would be compared with the performance data. This paper describes the neutronics, heat transfer, and cycle dynamics of this proposed system. Safety and radiation issues are presented. The views expressed in this document are those of the author and do not necessarily reflect agreement by the government.

  15. Soviet space nuclear reactor incidents - Perception versus reality

    NASA Technical Reports Server (NTRS)

    Bennett, Gary L.

    1992-01-01

    Since the Soviet Union reportedly began flying nuclear power sources in 1965 it has had four publicly known accidents involving space reactors, two publicly known accidents involving radioisotope power sources and one close call with a space reactor (Cosmos 1900). The reactor accidents, particularly Cosmos 954 and Cosmos 1402, indicated that the Soviets had adopted burnup as their reentry philosophy which is consistent with the U.S. philosophy from the 1960s and 1970s. While quantitative risk analyses have shown that the Soviet accidents have not posed a serious risk to the world's population, concerns still remain about Soviet space nuclear safety practices.

  16. Validation of High-Fidelity Reactor Physics Models for Support of the KJRR Experimental Campaign in the Advanced Test Reactor

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Nigg, David W.; Nielsen, Joseph W.; Norman, Daren R.

    The Korea Atomic Energy Research Institute is currently in the process of qualifying a Low-Enriched Uranium fuel element design for the new Ki-Jang Research Reactor (KJRR). As part of this effort, a prototype KJRR fuel element was irradiated for several operating cycles in the Northeast Flux Trap of the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The KJRR fuel element contained a very large quantity of fissile material (618g 235U) in comparison with historical ATR experiment standards (<1g 235U), and its presence in the ATR flux trap was expected to create a neutronic configuration that would be wellmore » outside of the approved validation envelope for the reactor physics analysis methods used to support ATR operations. Accordingly it was necessary, prior to high-power irradiation of the KJRR fuel element in the ATR, to conduct an extensive set of new low-power physics measurements with the KJRR fuel element installed in the ATR Critical Facility (ATRC), a companion facility to the ATR that is located in an immediately adjacent building, sharing the same fuel handling and storage canal. The new measurements had the objective of expanding the validation envelope for the computational reactor physics tools used to support ATR operations and safety analysis to include the planned KJRR irradiation in the ATR and similar experiments that are anticipated in the future. The computational and experimental results demonstrated that the neutronic behavior of the KJRR fuel element in the ATRC is well-understood, both in terms of its general effects on core excess reactivity and fission power distributions, its effects on the calibration of the core lobe power measurement system, as well as in terms of its own internal fission rate distribution and total fission power per unit ATRC core power. Taken as a whole, these results have significantly extended the ATR physics validation envelope, thereby enabling an entire new class of irradiation experiments.« less

  17. Integrating Safety Assessment Methods using the Risk Informed Safety Margins Characterization (RISMC) Approach

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Curtis Smith; Diego Mandelli

    Safety is central to the design, licensing, operation, and economics of nuclear power plants (NPPs). As the current light water reactor (LWR) NPPs age beyond 60 years, there are possibilities for increased frequency of systems, structures, and components (SSC) degradations or failures that initiate safety significant events, reduce existing accident mitigation capabilities, or create new failure modes. Plant designers commonly “over-design” portions of NPPs and provide robustness in the form of redundant and diverse engineered safety features to ensure that, even in the case of well-beyond design basis scenarios, public health and safety will be protected with a very highmore » degree of assurance. This form of defense-in-depth is a reasoned response to uncertainties and is often referred to generically as “safety margin.” Historically, specific safety margin provisions have been formulated primarily based on engineering judgment backed by a set of conservative engineering calculations. The ability to better characterize and quantify safety margin is important to improved decision making about LWR design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margin management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. In addition, as research and development (R&D) in the LWR Sustainability (LWRS) Program and other collaborative efforts yield new data, sensors, and improved scientific understanding of physical processes that govern the aging and degradation of plant SSCs needs and opportunities to better optimize plant safety and performance will become known. To support decision making related to economics, readability, and safety, the RISMC Pathway provides methods and tools that enable mitigation options known as margins management strategies. The purpose of the RISMC Pathway R&D is to support plant decisions for risk

  18. General aviation crash safety program at Langley Research Center

    NASA Technical Reports Server (NTRS)

    Thomson, R. G.

    1976-01-01

    The purpose of the crash safety program is to support development of the technology to define and demonstrate new structural concepts for improved crash safety and occupant survivability in general aviation aircraft. The program involves three basic areas of research: full-scale crash simulation testing, nonlinear structural analyses necessary to predict failure modes and collapse mechanisms of the vehicle, and evaluation of energy absorption concepts for specific component design. Both analytical and experimental methods are being used to develop expertise in these areas. Analyses include both simplified procedures for estimating energy absorption capabilities and more complex computer programs for analysis of general airframe response. Full-scale tests of typical structures as well as tests on structural components are being used to verify the analyses and to demonstrate improved design concepts.

  19. Helium-3 blankets for tritium breeding in fusion reactors

    NASA Technical Reports Server (NTRS)

    Steiner, Don; Embrechts, Mark; Varsamis, Georgios; Vesey, Roger; Gierszewski, Paul

    1988-01-01

    It is concluded that He-3 blankets offers considerable promise for tritium breeding in fusion reactors: good breeding potential, low operational risk, and attractive safety features. The availability of He-3 resources is the key issue for this concept. There is sufficient He-3 from decay of military stockpiles to meet the International Thermonuclear Experimental Reactor needs. Extraterrestrial sources of He-3 would be required for a fusion power economy.

  20. Make safety awareness a priority: Use a login software in your research facility

    DOE PAGES

    Camino, Fernando E.

    2017-01-21

    We report on a facility login software, whose objective is to improve safety in multi-user research facilities. Its most important safety features are: 1) blocks users from entering the lab after being absent for more than a predetermined number of days; 2) gives users a random safety quiz question, which they need to answer satisfactorily in order to use the facility; 3) blocks unauthorized users from using the facility afterhours; and 4) displays the current users in the facility. Besides restricting access to unauthorized users, the software keeps users mindful of key safety concepts. In addition, integration of the softwaremore » with a door controller system can convert it into an effective physical safety mechanism. Depending on DOE approval, the code may be available as open source.« less

  1. Make safety awareness a priority: Use a login software in your research facility

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Camino, Fernando E.

    We report on a facility login software, whose objective is to improve safety in multi-user research facilities. Its most important safety features are: 1) blocks users from entering the lab after being absent for more than a predetermined number of days; 2) gives users a random safety quiz question, which they need to answer satisfactorily in order to use the facility; 3) blocks unauthorized users from using the facility afterhours; and 4) displays the current users in the facility. Besides restricting access to unauthorized users, the software keeps users mindful of key safety concepts. In addition, integration of the softwaremore » with a door controller system can convert it into an effective physical safety mechanism. Depending on DOE approval, the code may be available as open source.« less

  2. DR Reactor VSR channel damage

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Kempf, F.J.; Rawlins, J.K.

    1961-10-30

    On July 11, 1961 the Ball 3X System at DR Reactor was inadventently tripped. All vertical safety rods dropped and all channels were filled with balls. This report has the twofold purpose of documenting borescope observations of ten vertical rod channels at DR Reactor and recording the estimated extent of graphite damage resulting from the above incident. Channel damage data are presented on appended drawings. With suitable notations, the tracings of these drawings may be revised to reflect any future graphite damage. All vertical rod channels at DR Reactor were visually examined with a closed circuit television system during ballmore » removal efforts. Typical photographs of trapped balls and ledges, as viewed on the television monitor, are shown. Photographs of typical graphite damage, obtained through the borescope are also included in this report. 3 refs., 8 figs., 1 tab.« less

  3. Developing a research agenda for patient safety in primary care. Background, aims and output of the LINNEAUS collaboration on patient safety in primary care.

    PubMed

    Esmail, Aneez; Valderas, Jose M; Verstappen, Wim; Godycki-Cwirko, Maciek; Wensing, Michel

    2015-09-01

    This paper is an introduction to a supplement to The European Journal of General Practice, bringing together a body of research focusing on the issue of patient safety in relation to primary care. The supplement represents the outputs of the LINNEAUS collaboration on patient safety in primary care, which was a four-year (2009-2013) coordination and support action funded under the Framework 7 programme by the European Union. Being a coordination and support action, its aim was not to undertake new research, but to build capacity through engaging primary care researchers and practitioners in identifying some of the key challenges in this area and developing consensus statements, which will be an essential part in developing a future research agenda. This introductory article describes the aims of the LINNEAUS collaboration, provides a brief summary of the reasons to focus on patient safety in primary care, the epidemiological and policy considerations, and an introduction to the papers included in the supplement.

  4. Developing a research agenda for patient safety in primary care. Background, aims and output of the LINNEAUS collaboration on patient safety in primary care

    PubMed Central

    Esmail, Aneez; Valderas, Jose M.; Verstappen, Wim; Godycki-Cwirko, Maciek; Wensing, Michel

    2015-01-01

    ABSTRACT This paper is an introduction to a supplement to The European Journal of General Practice, bringing together a body of research focusing on the issue of patient safety in relation to primary care. The supplement represents the outputs of the LINNEAUS collaboration on patient safety in primary care, which was a four-year (2009–2013) coordination and support action funded under the Framework 7 programme by the European Union. Being a coordination and support action, its aim was not to undertake new research, but to build capacity through engaging primary care researchers and practitioners in identifying some of the key challenges in this area and developing consensus statements, which will be an essential part in developing a future research agenda. This introductory article describes the aims of the LINNEAUS collaboration, provides a brief summary of the reasons to focus on patient safety in primary care, the epidemiological and policy considerations, and an introduction to the papers included in the supplement. PMID:26339828

  5. Oak Ridge National Laboratory Support of Non-light Water Reactor Technologies: Capabilities Assessment for NRC Near-term Implementation Action Plans for Non-light Water Reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Belles, Randy; Jain, Prashant K.; Powers, Jeffrey J.

    The Oak Ridge National Laboratory (ORNL) has a rich history of support for light water reactor (LWR) and non-LWR technologies. The ORNL history involves operation of 13 reactors at ORNL including the graphite reactor dating back to World War II, two aqueous homogeneous reactors, two molten salt reactors (MSRs), a fast-burst health physics reactor, and seven LWRs. Operation of the High Flux Isotope Reactor (HFIR) has been ongoing since 1965. Expertise exists amongst the ORNL staff to provide non-LWR training; support evaluation of non-LWR licensing and safety issues; perform modeling and simulation using advanced computational tools; run laboratory experiments usingmore » equipment such as the liquid salt component test facility; and perform in-depth fuel performance and thermal-hydraulic technology reviews using a vast suite of computer codes and tools. Summaries of this expertise are included in this paper.« less

  6. KERENA safety concept in the context of the Fukushima accident

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Zacharias, T.; Novotny, C.; Bielor, E.

    Within the last three years AREVA NP and E.On KK finalized the basic design of KERENA which is a medium sized innovative boiling water reactor, based on the operational experience of German BWR nuclear power plants (NPPs). It is a generation III reactor design with a net electrical output of about 1250 MW. It combines active safety equipment of service-proven designs with new passive safety components, both safety classified. The passive systems utilize basic laws of physics, such as gravity and natural convection, enabling them to function without electric power. Even actuation of these systems is performed thanks to basicmore » physic laws. The degree of diversity in component and system design, achieved by combining active and passive equipment, results in a very low core damage frequency. The Fukushima accident enhanced the world wide discussion about the safety of operating nuclear power plants. World wide stress tests for operating nuclear power plants are being performed embracing both natural and man made hazards. Beside the assessment of existing power plants, also new designs are analyzed regarding the system response to beyond design base accidents. KERENA's optimal combination of diversified cooling systems (active and passive) allows passing efficiently such tests, with a high level of confidence. This paper describes the passive safety components and the KERENA reactor behavior after a Fukushima like accident. (authors)« less

  7. National Safety Council

    MedlinePlus

    ... Safety Management Systems Workplace Safety Consulting Employee Perception Surveys Research Journey to Safety Excellence Join the Journey What ... Safety Management Systems Workplace Safety Consulting Employee Perception Surveys Research Journey to Safety Excellence Join the Journey What ...

  8. Autonomous Control of Space Reactor Systems

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Belle R. Upadhyaya; K. Zhao; S.R.P. Perillo

    2007-11-30

    Autonomous and semi-autonomous control is a key element of space reactor design in order to meet the mission requirements of safety, reliability, survivability, and life expectancy. Interrestrial nuclear power plants, human operators are avilable to perform intelligent control functions that are necessary for both normal and abnormal operational conditions.

  9. MTR, TRA603. FIRST FLOOR PLAN. REACTOR AT CENTER. TWENTYMETER CHOPPER ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    MTR, TRA-603. FIRST FLOOR PLAN. REACTOR AT CENTER. TWENTY-METER CHOPPER HOUSE. COFFIN TURNING ROLLS. REMOVABLE PANEL OVER CANAL ON EAST SIDE. NEW PLUG STORAGE ACCESS. DOOR SCHEDULE INDICATES STEEL (FOR VAULT), WIRE MESH, AND HOLLOW METAL TYPES. STORAGE AND ISSUE ROOM. SAFETY SHOWERS. DOORWAY TO WING, TRA-604. BLAW-KNOX 3150-803-2, 7/1950. INL INDEX NO. 531-0603-00-098-100561, REV. 10. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  10. Safety monitoring and reactor transient interpreter

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Hench, J. E.; Fukushima, T. Y.

    1983-12-20

    An apparatus which monitors a subset of control panel inputs in a nuclear reactor power plant, the subset being those indicators of plant status which are of a critical nature during an unusual event. A display (10) is provided for displaying primary information (14) as to whether the core is covered and likely to remain covered, including information as to the status of subsystems needed to cool the core and maintain core integrity. Secondary display information (18,20) is provided which can be viewed selectively for more detailed information when an abnormal condition occurs. The primary display information has messages (24)more » for prompting an operator as to which one of a number of pushbuttons (16) to press to bring up the appropriate secondary display (18,20). The apparatus utilizes a thermal-hydraulic analysis to more accurately determine key parameters (such as water level) from other measured parameters, such as power, pressure, and flow rate.« less

  11. RELAP5 Analysis of the Hybrid Loop-Pool Design for Sodium Cooled Fast Reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Hongbin Zhang; Haihua Zhao; Cliff Davis

    2008-06-01

    An innovative hybrid loop-pool design for sodium cooled fast reactors (SFR-Hybrid) has been recently proposed. This design takes advantage of the inherent safety of a pool design and the compactness of a loop design to improve economics and safety of SFRs. In the hybrid loop-pool design, primary loops are formed by connecting the reactor outlet plenum (hot pool), intermediate heat exchangers (IHX), primary pumps and the reactor inlet plenum with pipes. The primary loops are immersed in the cold pool (buffer pool). Passive safety systems -- modular Pool Reactor Auxiliary Cooling Systems (PRACS) – are added to transfer decay heatmore » from the primary system to the buffer pool during loss of forced circulation (LOFC) transients. The primary systems and the buffer pool are thermally coupled by the PRACS, which is composed of PRACS heat exchangers (PHX), fluidic diodes and connecting pipes. Fluidic diodes are simple, passive devices that provide large flow resistance in one direction and small flow resistance in reverse direction. Direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) are immersed in the cold pool to transfer decay heat to the environment by natural circulation. To prove the design concepts, especially how the passive safety systems behave during transients such as LOFC with scram, a RELAP5-3D model for the hybrid loop-pool design was developed. The simulations were done for both steady-state and transient conditions. This paper presents the details of RELAP5-3D analysis as well as the calculated thermal response during LOFC with scram. The 250 MW thermal power conventional pool type design of GNEP’s Advanced Burner Test Reactor (ABTR) developed by Argonne National Laboratory was used as the reference reactor core and primary loop design. The reactor inlet temperature is 355 °C and the outlet temperature is 510 °C. The core design is the same as that for ABTR. The steady state buffer pool temperature is the same as the reactor

  12. Top shield temperatures, C and K Reactors

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    Agar, J.D.

    1964-12-28

    A modification program is now in progress at the C and K Reactors consisting of an extensive renovation of the graphite channels in the vertical safety rod ststems. The present VSR channels are being enlarged by a graphite coring operation and channel sleeves will be installed in the larger channels. One problem associated with the coring operation is the danger of damaging top thermal shield cooling tubes located close to the VSR channels to such an extent that these tubes will have to be removed from service. If such a condition should exist at one or a number of locationsmore » in the top shield of the reactors after reactor startup, the question remains -- what would the resulting temperatures be of the various components of the top shields? This study was initiated to determine temperature distributions in the top shield complex at the C and K Reactors for various top thermal shield coolant system conditions. Since the top thermal shield cooling system at C Reactor is different than those at the K Reactors, the study was conducted separately for the two different systems.« less

  13. Labor unions and safety climate: perceived union safety values and retail employee safety outcomes.

    PubMed

    Sinclair, Robert R; Martin, James E; Sears, Lindsay E

    2010-09-01

    Although trade unions have long been recognized as a critical advocate for employee safety and health, safety climate research has not paid much attention to the role unions play in workplace safety. We proposed a multiple constituency model of workplace safety which focused on three central safety stakeholders: top management, ones' immediate supervisor, and the labor union. Safety climate research focuses on management and supervisors as key stakeholders, but has not considered whether employee perceptions about the priority their union places on safety contributes contribute to safety outcomes. We addressed this gap in the literature by investigating unionized retail employee (N=535) perceptions about the extent to which their top management, immediate supervisors, and union valued safety. Confirmatory factor analyses demonstrated that perceived union safety values could be distinguished from measures of safety training, workplace hazards, top management safety values, and supervisor values. Structural equation analyses indicated that union safety values influenced safety outcomes through its association with higher safety motivation, showing a similar effect as that of supervisor safety values. These findings highlight the need for further attention to union-focused measures related to workplace safety as well as further study of retail employees in general. We discuss the practical implications of our findings and identify several directions for future safety research. 2009 Elsevier Ltd. All rights reserved.

  14. SAFETY AND SECURITY BUILDING, TRA614. ELEVATIONS. SECTIONS. TWO ROOF LEVELS. ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    SAFETY AND SECURITY BUILDING, TRA-614. ELEVATIONS. SECTIONS. TWO ROOF LEVELS. BLAW-KNOX 3150-814-2, 3/1950. INL INDEX NO. 531-0614-00-098-100703, REV. 6. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  15. Development concept for a small, split-core, heat-pipe-cooled nuclear reactor

    NASA Technical Reports Server (NTRS)

    Lantz, E.; Breitwieser, R.; Niederauer, G. F.

    1974-01-01

    There have been two main deterrents to the development of semiportable nuclear reactors. One is the high development costs; the other is the inability to satisfy with assurance the questions of operational safety. This report shows how a split-core, heat-pipe cooled reactor could conceptually eliminate these deterrents, and examines and summarizes recent work on split-core, heat-pipe reactors. A concept for a small reactor that could be developed at a comparatively low cost is presented. The concept would extend the technology of subcritical radioisotope thermoelectric generators using 238 PuO2 to the evolution of critical space power reactors using 239 PuO2.

  16. The Union RAP: Industry-Wide Research-Action Projects to Win Health and Safety Improvements

    PubMed Central

    McQuiston, Thomas H.; Lippin, Tobi Mae; Anderson, Leeann G.; Beach, M. Josie; Frederick, James; Seymour, Thomas A.

    2009-01-01

    Unions are ripe to engage in community-based participatory research (CBPR). We briefly profile 3 United Steelworker CBPR projects aimed at uncovering often-undocumented, industry-wide health and safety conditions in which US industrial workers toil. The results are to be used to advocate improvements at workplace, industry, and national policy levels. We offer details of our CBPR approach (Research-Action Project [RAP]) that engages workers and others in all research stages. Elements of RAPs include strategically constructed teams with knowledge of the industry and health and safety and with skills in research, participatory facilitation, and training; reciprocal training on these knowledge and skill areas; iterative processes of large and small group work; use of technology; and facilitator-developed tools and intermediate products. PMID:19890145

  17. Characterisation of the epithermal neutron irradiation facility at the Portuguese research reactor using MCNP.

    PubMed

    Beasley, D G; Fernandes, A C; Santos, J P; Ramos, A R; Marques, J G; King, A

    2015-05-01

    The radiation field at the epithermal beamline and irradiation chamber installed at the Portuguese Research Reactor (RPI) at the Campus Tecnológico e Nuclear of Instituto Superior Técnico was characterised in the context of Prompt Gamma Neutron Activation Analysis (PGNAA) applications. Radiographic films, activation foils and thermoluminescence dosimeters were used to measure the neutron fluence and photon dose rates in the irradiation chamber. A fixed-source MCNPX model of the beamline and chamber was developed and compared to measurements in the first step towards planning a new irradiation chamber. The high photon background from the reactor results in the saturation of the detector and the current facility configuration yields an intrinsic insensitivity to various elements of interest for PGNAA. These will be addressed in future developments. Copyright © 2015 Elsevier Ltd. All rights reserved.

  18. Investigation of materials for fusion power reactors

    NASA Astrophysics Data System (ADS)

    Bouhaddane, A.; Slugeň, V.; Sojak, S.; Veterníková, J.; Petriska, M.; Bartošová, I.

    2014-06-01

    The possibility of application of nuclear-physical methods to observe radiation damage to structural materials of nuclear facilities is nowadays a very actual topic. The radiation damage to materials of advanced nuclear facilities, caused by extreme radiation stress, is a process, which significantly limits their operational life as well as their safety. In the centre of our interest is the study of the radiation degradation and activation of the metals and alloys for the new nuclear facilities (Generation IV fission reactors, fusion reactors ITER and DEMO). The observation of the microstructure changes in the reactor steels is based on experimental investigation using the method of positron annihilation spectroscopy (PAS). The experimental part of the work contains measurements focused on model reactor alloys and ODS steels. There were 12 model reactor steels and 3 ODS steels. We were investigating the influence of chemical composition on the production of defects in crystal lattice. With application of the LT 9 program, the spectra of specimen have been evaluated and the most convenient samples have been determined.

  19. The neutron texture diffractometer at the China Advanced Research Reactor

    NASA Astrophysics Data System (ADS)

    Li, Mei-Juan; Liu, Xiao-Long; Liu, Yun-Tao; Tian, Geng-Fang; Gao, Jian-Bo; Yu, Zhou-Xiang; Li, Yu-Qing; Wu, Li-Qi; Yang, Lin-Feng; Sun, Kai; Wang, Hong-Li; Santisteban, J. r.; Chen, Dong-Feng

    2016-03-01

    The first neutron texture diffractometer in China has been built at the China Advanced Research Reactor, due to strong demand for texture measurement with neutrons from the domestic user community. This neutron texture diffractometer has high neutron intensity, moderate resolution and is mainly applied to study texture in commonly used industrial materials and engineering components. In this paper, the design and characteristics of this instrument are described. The results for calibration with neutrons and quantitative texture analysis of zirconium alloy plate are presented. The comparison of texture measurements with the results obtained in HIPPO at LANSCE and Kowari at ANSTO illustrates the reliability of the texture diffractometer. Supported by National Nature Science Foundation of China (11105231, 11205248, 51327902) and International Atomic Energy Agency-TC program (CPR0012)

  20. Overview of the 2014 Edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook)

    DOE Office of Scientific and Technical Information (OSTI.GOV)

    John D. Bess; J. Blair Briggs; Jim Gulliford

    2014-10-01

    The International Reactor Physics Experiment Evaluation Project (IRPhEP) is a widely recognized world class program. The work of the IRPhEP is documented in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Integral data from the IRPhEP Handbook is used by reactor safety and design, nuclear data, criticality safety, and analytical methods development specialists, worldwide, to perform necessary validations of their calculational techniques. The IRPhEP Handbook is among the most frequently quoted reference in the nuclear industry and is expected to be a valuable resource for future decades.