Science.gov

Sample records for reactor safety systems

  1. Reactor safety assessment system

    SciTech Connect

    Sebo, D.E.; Bray, M.A.; King, M.A.

    1987-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSA is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category.

  2. 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.

  3. Safety program considerations for space nuclear reactor systems

    SciTech Connect

    Cropp, L.O.

    1984-08-01

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

  4. Software reliability and safety in nuclear reactor protection systems

    SciTech Connect

    Lawrence, J.D.

    1993-11-01

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

  5. 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. PMID:18049233

  6. Simulation of the modified K reactor supplementary safety system

    SciTech Connect

    Paik, I.K.; Canas, L.R. ); Peterson, P.F. )

    1991-01-01

    The supplementary safety system (SSS) of the K reactor provides a second line of defense to shut down the reactor if the safety and control rods fail to scram. The SSS was originally designed to inject a neutron poison solution (ink) into the reactor tank via spargers. Recently, concerns arose that the ink inventory might run out before the ink front returned to the moderator during a loss-of-ac-power transient in which the coolant pumps coast down. Thus, a new system has been added to inject additional ink through the pump suctions so that ink will arrive in the core before depletion of the sparger ink. The MODFLOW code was developed to calculate the moderator flow distribution in Savannah River site (SRS) reactors, including the effects of inertia and stratification from buoyancy forces.

  7. Nuclear safety as applied to space power reactor systems

    SciTech Connect

    Cummings, G.E.

    1987-01-01

    Current space nuclear power reactor safety issues are discussed with respect to the unique characteristics of these reactors. An approach to achieving adequate safety and a perception of safety is outlined. This approach calls for a carefully conceived safety program which makes uses of lessons learned from previous terrestrial power reactor development programs. This approach includes use of risk analyses, passive safety design features, and analyses/experiments to understand and control off-design conditions. The point is made that some recent accidents concerning terrestrial power reactors do not imply that space power reactors cannot be operated safety.

  8. Nuclear safety as applied to space power reactor systems

    SciTech Connect

    Cummings, G.E.

    1987-01-01

    To develop a strategy for incorporating and demonstrating safety, it is necessary to enumerate the unique aspects of space power reactor systems from a safety standpoint. These features must be differentiated from terrestrial nuclear power plants so that our experience can be applied properly. Some ideas can then be developed on how safe designs can be achieved so that they are safe and perceived to be safe by the public. These ideas include operating only after achieving a stable orbit, developing an inherently safe design, ''designing'' in safety from the start and managing the system development (design) so that it is perceived safe. These and other ideas are explored further in this paper.

  9. PWR systems transient analysis: a reactor-safety perspective

    SciTech Connect

    Kennedy, M.F.; Abramson, P.B.; McDonald, T.A.

    1982-01-01

    In the simulation of transient events in large PWR reactor systems for reactor safety studies, the plant model is quite detailed and must include most of the plant components and control systems to adequately analyze the range of transients. The results discussed were calculated with the RELAP4/MOD6 code and reveal the need for the analysis to carefully review and understand the results to assure that they are not being adversely affected by the improper solution techniques or changes in models during the calculation.

  10. Passive compact molten salt reactor (PCMSR), modular thermal breeder reactor with totally passive safety system

    NASA Astrophysics Data System (ADS)

    Harto, Andang Widi

    2012-06-01

    Design Study Passive Compact Molten Salt Reactor (PCMSR) with totally passive safety system has been performed. The term of Compact in the PCMSR name means that the reactor system is designed to have relatively small volume per unit power output by using modular and integral concept. In term of modular, the reactor system consists of three modules, i.e. reactor module, turbine module and fuel management module. The reactor module is an integral design that consists of reactor, primary and intermediate heat exchangers and passive post shutdown cooling system. The turbine module is an integral design of a multi heating, multi cooling, regenerative gas turbine. The fuel management module consists of all equipments related to fuel preparation, fuel reprocessing and radioactive handling. The preliminary calculations show that the PCMSR has negative temperature and void reactivity coefficient, passive shutdown characteristic related to fuel pump failure and possibility of using natural circulation for post shutdown cooling system.

  11. Passive compact molten salt reactor (PCMSR), modular thermal breeder reactor with totally passive safety system

    SciTech Connect

    Harto, Andang Widi

    2012-06-06

    Design Study Passive Compact Molten Salt Reactor (PCMSR) with totally passive safety system has been performed. The term of Compact in the PCMSR name means that the reactor system is designed to have relatively small volume per unit power output by using modular and integral concept. In term of modular, the reactor system consists of three modules, i.e. reactor module, turbine module and fuel management module. The reactor module is an integral design that consists of reactor, primary and intermediate heat exchangers and passive post shutdown cooling system. The turbine module is an integral design of a multi heating, multi cooling, regenerative gas turbine. The fuel management module consists of all equipments related to fuel preparation, fuel reprocessing and radioactive handling. The preliminary calculations show that the PCMSR has negative temperature and void reactivity coefficient, passive shutdown characteristic related to fuel pump failure and possibility of using natural circulation for post shutdown cooling system.

  12. Safety aspects of Particle Bed Reactor plutonium burner system

    SciTech Connect

    Powell, J.R.; Ludewig, H.; Todosow, M.

    1993-08-01

    An assessment is made of the safety aspects peculiar to using the Particle Bed Reactor (PBR) as the burner in a plutonium disposal system. It is found that a combination of the graphitic fuel, high power density possible with the PBR and engineered design features results in an attractive concept. The high power density potentially makes it possible to complete the plutonium burning without requiring reprocessing and remanufacturing fuel. This possibility removes two hazardous steps from a plutonium burning complex. Finally, two backup cooling systems depending on thermo-electric converters and heat pipes act as ultimate heat removal sinks in the event of accident scenarios which result in loss of fuel cooling.

  13. Safety characteristics of a suspended-pellet fission reactor system

    NASA Astrophysics Data System (ADS)

    Kingdon, David Ross

    A new fission reactor system with passive safety characteristics to eliminate the occurrence of loss-of-coolant accidents, reduce reactivity excursion effects, and which also provides for closure of the nuclear fuel cycle through on-site spent fuel management is examined. The concept uses multi-coated fuel pellets which are suspended by an upward moving coolant in vertical columns of the reactor core and electro-refining elemental separation to remove selected fission products prior to actinide recycling. The possibility of fuel melt following a loss-of-coolant is avoided as a decrease in coolant flow results in the removal of fuel from the core through the action of gravity alone. Average fluid velocities in the columns which are necessary to suspend the pellets are calculated and found to be consistent with the necessary heat extraction to yield ˜1--10 Wth per column. The total output power of such suspended pellet-type reactors is compared to the power necessary to provide the suspending fluid flow, yielding favourable ratios of ˜102--103. The reduction of reactivity excursion tendencies is envisaged through an ablative layer of material in the pellets which sublimates at temperatures above normal operating conditions. In the event of a power or temperature increase the particles fragment and thereby change their hydrodynamic drag characteristics, thus leading to fuel removal from the core by elutriation. Comparison of nuclear-to-thermal response times and elutriation rates for limiting power transients indicate that the present design assists in reactivity excursion mitigation. Closure of the nuclear fuel cycle is attained through a spent fuel management strategy which requires only on-site storage of a fraction of the fission products produced during reactor operation. Electro-refining separation of selected fission products combined with complete actinide recycling yields no isolation of plutonium or highly enriched uranium during the procedure. The out

  14. 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.

  15. Identification and characterization of passive safety system and inherent safety feature building blocks for advanced light-water reactors

    SciTech Connect

    Forsberg, C.W.

    1989-01-01

    Oak Ridge National Laboratory (ORNL) is investigating passive and inherent safety options for Advanced Light-Water Reactors (ALWRs). A major activity in 1989 includes identification and characterization of passive safety system and inherent safety feature building blocks, both existing and proposed, for ALWRs. Preliminary results of this work are reported herein. This activity is part of a larger effort by the US Department of Energy, reactor vendors, utilities, and others in the United States to develop improved LWRs. The Advanced Boiling Water Reactor (ABWR) program and the Advanced Pressurized Water Reactor (APWR) program have as goals improved, commercially available LWRs in the early 1990s. The Advanced Simplified Boiling Water Reactor (ASBWR) program and the AP-600 program are developing more advanced reactors with increased use of passive safety systems. It is planned that these reactors will become commercially available in the mid 1990s. The ORNL program is an exploratory research program for LWRs beyond the year 2000. Desired long-term goals for such reactors include: (1) use of only passive and inherent safety, (2) foolproof against operator errors, (3) malevolence resistance against internal sabotage and external assault and (4) walkaway safety. The acronym ''PRIME'' (Passive safety, Resilient operation, Inherent safety, Malevolence resistance, and Extended (walkaway) safety) is used to summarize these desired characteristics. Existing passive and inherent safety options are discussed in this document.

  16. Reactor operation safety information document

    SciTech Connect

    Not Available

    1990-01-01

    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  17. Reactor Safety Research Programs

    SciTech Connect

    Edler, S. K.

    1981-07-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  18. Plant Modernization with Digital Reactor Protection System Safety System Upgrades at US Nuclear Power Stations

    SciTech Connect

    Heckle, Wm. Lloyd; Bolian, Tricia W.

    2006-07-01

    As the current fleet of nuclear power plants in the US reaches 25+ years of operation, obsolescence is driving many utilities to implement upgrades to both their safety and non-safety-related Instrumentation and Control (I and C) Systems. Digital technology is the predominant replacement technology for these upgrades. Within the last 15 years, digital control systems have been deployed in non-safety- related control applications at many utilities. In addition, a few utilities have replaced small safety-related systems utilizing digital technology. These systems have shown digital technology to be robust, reliable and simpler to maintain. Based upon this success, acceptance of digital technology has gained momentum with both utilities and regulatory agencies. Today, in an effort to extend the operating lives of their nuclear stations and resolve obsolescence of critical components, utilities are now pursuing digital technology for replacement of their primary safety systems. AREVA is leading this effort in the United States with the first significant digital upgrade of a major safety system. AREVA has previously completed upgrades to safety-related control systems emergency diesel engine controls and governor control systems for a hydro station which serves as the emergency power source for a nuclear station. Currently, AREVA is implementing the replacement of both the Reactor Protection System (RPS) and the Engineered Safety Features Actuation System (ESFAS) on all three units at a US PWR site. (authors)

  19. Nuclear Reactor Safety: a current awareness bulletin

    SciTech Connect

    Cunningham, D.C.

    1985-01-15

    Nuclear Reactor Safety announces on a semimonthly basis the current worldwide information available on all safety-related aspects of fission reactors, including: accident analysis, safety systems, radiation protection, decommissioning and dismantling, and security measures.

  20. Conceptual Design of Passive Safety System for Lead-Bismuth Cooled Fast Reactor

    NASA Astrophysics Data System (ADS)

    Abdullah, A. G.; Nandiyanto, A. B. D.

    2016-04-01

    This paper presents the results of the conceptual design of passive safety systems for reactor power 225 MWth using Pb-Bi coolant. Main purpose of this research is to design of heat removal system from the reactor wall. The heat from the reactor wall is removed by RVACS system using the natural circulation from the atmosphere around the reactor at steady state. The calculation is performed numerically using Newton-Raphson method. The analysis involves the heat transfer systems in a radiation, conduction and natural convection. Heat transfer calculations is performed on the elements of the reactor vessel, outer wall of guard vessel and the separator plate. The simulation results conclude that the conceptual design is able to remove heat 1.33% to 4.67% from the thermal reactor power. It’s can be hypothesized if the reactor had an accident, the system can still overcome the heat due to decay.

  1. 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.

  2. Reactor Safety Research Programs

    SciTech Connect

    Dotson, CW

    1980-08-01

    This document summarizes the work performed by Pacific Northwest laboratory from October 1 through December 31, 1979, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, lspra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  3. 77 FR 62270 - Proposed Revision Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-12

    ... COMMISSION Proposed Revision Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors AGENCY... Treatment of Non-Safety Systems (RTNSS) for Passive Advanced Light Water Reactors.'' The current SRP does not contain guidance on the proposed RTNSS for Passive Advance Light Water Reactors. DATES:...

  4. Reviewing real-time performance of nuclear reactor safety systems

    SciTech Connect

    Preckshot, G.G.

    1993-08-01

    The purpose of this paper is to recommend regulatory guidance for reviewers examining real-time performance of computer-based safety systems used in nuclear power plants. Three areas of guidance are covered in this report. The first area covers how to determine if, when, and what prototypes should be required of developers to make a convincing demonstration that specific problems have been solved or that performance goals have been met. The second area has recommendations for timing analyses that will prove that the real-time system will meet its safety-imposed deadlines. The third area has description of means for assessing expected or actual real-time performance before, during, and after development is completed. To ensure that the delivered real-time software product meets performance goals, the paper recommends certain types of code-execution and communications scheduling. Technical background is provided in the appendix on methods of timing analysis, scheduling real-time computations, prototyping, real-time software development approaches, modeling and measurement, and real-time operating systems.

  5. Testing of Passive Safety System Performance for Higher Power Advanced Reactors

    SciTech Connect

    brian G. Woods; Jose Reyes, Jr.; John Woods; John Groome; Richard Wright

    2004-12-31

    This report describes the results of NERI research on the testing of advanced passive safety performance for the Westinghouse AP1000 design. The objectives of this research were: (a) to assess the AP1000 passive safety system core cooling performance under high decay power conditions for a spectrum of breaks located at a variety of locations, (b) to compare advanced thermal hydraulic computer code predictions to the APEX high decay power test data and (c) to develop new passive safety system concepts that could be used for Generation IV higher power reactors.

  6. A concept of JAERI passive safety light water reactor system (JPSR)

    SciTech Connect

    Murao, Y.; Araya, F.; Iwamura, T.

    1995-09-01

    The Japan Atomic Energy Research Institute (JAERI) proposed a passive safety reactor system concept, JPSR, which was developed for reducing manpower in operation and maintenance and influence of human errors on reactor safety. In the concept the system was extremely simplified. The inherent matching nature of core generation and heat removal rate within a small volume change of the primary coolant is introduced by eliminating chemical shim and adopting in-vessel control rod drive mechanism units, a low power density core and once-through steam generators. In order to simplify the system, a large pressurizer, canned pumps, passive engineered-safety-features-system (residual heat removal system and coolant injection system) are adopted and the total system can be significantly simplified. The residual heat removal system is completely passively actuated in non-LOCAs and is also used for depressurization of the primary coolant system to actuate accumulators in small break LOCAs and reactor shutdown cooling system in normal operation. All of systems for nuclear steam supply system are built in the containment except for the air coolers as a the final heat sink of the passive residual heat removal system. Accordingly the reliability of the safety system and the normal operation system is improved, since most of residual heat removal system is always working and a heat sink for normal operation system is {open_quotes}safety class{close_quotes}. In the passive coolant injection system, depressurization of the primary cooling system by residual heat removal system initiates injection from accumulators designed for the MS-600 in medium pressure and initiates injection from the gravity driven coolant injection pool at low pressure. Analysis with RETRAN-02/MOD3 code demonstrated the capability of passive load-following, self-power-controllability, cooling and depressurization.

  7. 78 FR 41436 - Proposed Revision to Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-07-10

    ... solicitation for public comment published in the Federal Register on October 12, 2012 (77 FR 62270), on the... COMMISSION Proposed Revision to Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors... Treatment of Non-Safety Systems (RTNSS) for Passive Advanced Light Water Reactors.'' The NRC seeks...

  8. Enhancing VHTR passive safety and economy with thermal radiation based direct reactor auxiliary cooling system

    SciTech Connect

    Zhao, H.; Zhang, H.; Zou, L.; Sun, X.

    2012-07-01

    One of the most important requirements for Gen. IV Very High Temperature Reactor (VHTR) is passive safety. Currently all the gas cooled version of VHTR designs use Reactor Vessel Auxiliary Cooling System (RVACS) for passive decay heat removal. The RVACS can be characterized as a surface-based decay heat removal system. It is especially suitable for smaller power reactors since small systems have relatively larger surface area to volume ratio. However, RVACS limits the maximum achievable power level for modular VHTRs due to the mismatch between the reactor power (proportional to the core volume) and decay heat removal capability (proportional to the vessel surface area). Besides the safety considerations, VHTRs also need to be economical in order to compete with other reactor concepts and other types of energy sources. The limit of decay heat removal capability set by using RVACS has affected the economy of VHTRs. A potential alternative solution is to use a volume-based passive decay heat removal system, called Direct Reactor Auxiliary Cooling Systems (DRACS), to remove or mitigate the limitation on decay heat removal capability. DRACS composes of natural circulation loops with two sets of heat exchangers, one on the reactor side and another on the environmental side. For the reactor side, cooling pipes will be inserted into holes made in the outer or inner graphite reflector blocks. There will be gaps or annular regions formed between these cooling pipes and their corresponding surrounding graphite surfaces. Graphite has an excellent heat conduction property. By taking advantage of this feature, we can have a volume-based method to remove decay heat. The scalability can be achieved, if needed, by employing more rows of cooling pipes to accommodate higher decay heat rates. Since heat can easily conduct through the graphite regions among the holes made for the cooling pipes, those cooling pipes located further away from the active core region can still be very

  9. 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.

  10. PACTEL experiments for the investigation of passive safety injection systems of advanced light water reactors

    SciTech Connect

    Tuunanen, J.; Munther, R.; Vihavainen, J.

    1996-07-01

    An important aspect of Advanced Light Water Reactors (ALWRs) decay heat removal concerns the plant response under Loss of Coolant Accident (LOCA) conditions. In ALWRs, e.g. Westinghouse AP600, pump driven Emergency Core Cooling Systems (ECCS) are replaced by passive safety injection systems, which are gravity driven. It is therefore important that in such accidents, the ALWR coolant system pressure can be controlled to allow gravity fed injection to take place. The safety issue here is whether undesirable system responses could occur in any circumstances. Additionally, it is necessary to prove that the plant always depressurizes sufficiently for the ECCS to operate efficiently. Two experimental series have been carried out with the PACTEL integral test facility on the simulation of passive safety injection systems of ALWRs in accidental conditions. The safety system investigated was a passive core make-up tank (CMT), which was connected to the downcomer of the test facility. This paper starts with a short description of the PACTEL test facility and a summary of the results of the passive safety injection tests on PACTEL. Also included is a summary of the results of the computer simulations of the tests. The second part of the paper consists of a description of the planned third passive safety injection test series and the results of the pre-test simulations of the planned tests.

  11. Safety Design Strategy for the Advanced Test Reactor Emergency Firewater Injection System Replacement Project

    SciTech Connect

    Noel Duckwitz

    2011-06-01

    In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

  12. Enhancing VHTR Passive Safety and Economy with Thermal Radiation Based Direct Reactor Auxiliary Cooling System

    SciTech Connect

    Haihua Zhao; Hongbin Zhang; Ling Zou; Xiaodong Sun

    2012-06-01

    One of the most important requirements for Gen. IV Very High Temperature Reactor (VHTR) is passive safety. Currently all the gas cooled version of VHTR designs use Reactor Vessel Auxiliary Cooling System (RVACS) for passive decay heat removal. The decay heat first is transferred to the core barrel by conduction and radiation, and then to the reactor vessel by thermal radiation and convection; finally the decay heat is transferred to natural circulated air or water systems. RVACS can be characterized as a surface based decay heat removal system. The RVACS is especially suitable for smaller power reactors since small systems have relatively larger surface area to volume ratio. However, RVACS limits the maximum achievable power level for modular VHTRs due to the mismatch between the reactor power (proportional to volume) and decay heat removal capability (proportional to surface area). When the relative decay heat removal capability decreases, the peak fuel temperature increases, even close to the design limit. Annular core designs with inner graphite reflector can mitigate this effect; therefore can further increase the reactor power. Another way to increase the reactor power is to increase power density. However, the reactor power is also limited by the decay heat removal capability. Besides the safety considerations, VHTRs also need to be economical in order to compete with other reactor concepts and other types of energy sources. The limit of decay heat removal capability set by using RVACS has affected the economy of VHTRs. A potential alternative solution is to use a volume-based passive decay heat removal system, called Direct Reactor Auxiliary Cooling Systems (DRACS), to remove or mitigate the limitation on decay heat removal capability. DRACS composes of natural circulation loops with two sets of heat exchangers, one on the reactor side and another on the environment side. For the reactor side, cooling pipes will be inserted into holes made in the outer or

  13. Radiation safety assessment of a system of small reactors for distributed energy.

    PubMed

    Odano, N; Ishida, T

    2005-01-01

    A passively safe small reactor for a distributed energy system, PSRD, is an integral type of light-water reactor with a thermal output of 100 or 300 MW aimed to be used for supplying district heat, electricity to small grids, and so on. Candidate locations for the PSRD as a distributed energy source are on-ground, deep underground, and in a seaside pit in the vicinity of the energy consumption area. Assessments of the radiation safety of a PSRD were carried out for three cases corresponding to normal operation, shutdown and a hypothetical postulated accident for several siting candidates. Results of the radiation safety assessment indicate that the PSRD design has sufficient shielding performance and capability and that the exposure to the general public is very low in the case of a hypothetical accident. PMID:16381690

  14. Testing of the Multi-Application Small Light Water Reactor (MASLWR) Passive Safety Systems

    SciTech Connect

    Reyes, Jose N.; Groome, John; Woods, Brian G.; Young, Eric; Abel, Kent; Yao, You; Yeon Jong Yoo

    2006-07-01

    Experimental thermal hydraulic research has been conducted at Oregon State University for the purpose of assessing the performance of a new reactor design concept, the Multi-application Small Light Water Reactor (MASLWR). MASLWR is a pressurized light water reactor that uses natural circulation in both normal and transient operation. The purpose of the OSU MASLWR Test Facility is to assess the operation of the MASLWR under normal full pressure and full temperature conditions and to assess the passive safety systems under transient conditions. The data generated by the testing program will be used to assess computer code calculations and to provide a better understanding of the thermal-hydraulic phenomena in the design of the MASLWR NSSS. During this testing program, four tests were conducted at the OSU MASLWR Test Facility. These tests included one design basis accident and one beyond design basis accident. Plant start up, normal operation and shut down evolutions were also examined. (authors)

  15. Perspectives on reactor safety

    SciTech Connect

    Haskin, F.E.; Camp, A.L.

    1994-03-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  16. Part I. Fuel-motion diagnostics in support of fast-reactor safety experiments. Part II. Fission product detection system in support of fast reactor safety experiments

    SciTech Connect

    Devolpi, A.; Doerner, R.C.; Fink, C.L.; Regis, J.P.; Rhodes, E.A.; Stanford, G.S.; Braid, T.H.; Boyar, R.E.

    1986-05-01

    In all destructive fast-reactor safety experiments at TREAT, fuel motion and cladding failure have been monitored by the fast-neutron/gamma-ray hodoscope, providing experimental results that are directly applicable to design, modeling, and validation in fast-reactor safety. Hodoscope contributions to the safety program can be considered to fall into several groupings: pre-failure fuel motion, cladding failure, post-failure fuel motion, steel blockages, pretest and posttest radiography, axial-power-profile variations, and power-coupling monitoring. High-quality results in fuel motion have been achieved, and motion sequences have been reconstructed in qualitative and quantitative visual forms. A collimated detection system has been used to observe fission products in the upper regions of a test loop in the TREAT reactor. Particular regions of the loop are targeted through any of five channels in a rotatable assembly in a horizontal hole through the biological shield. A well-type neutron detector, optimized for delayed neutrons, and two GeLi gamma ray spectrometers have been used in several experiments. Data are presented showing a time history of the transport of Dn emitters, of gamma spectra identifying volatile fission products deposited as aerosols, and of fission gas isotopes released from the coolant.

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

    NASA Astrophysics Data System (ADS)

    Galvez, Cristhian

    2011-12-01

    The Pebble Bed Advanced High Temperature Reactor (PB-AHTR) is a pebble fueled, liquid salt cooled, high temperature nuclear reactor design that can be used for electricity generation or other applications requiring the availability of heat at elevated temperatures. A stage in the design evolution of this plant requires the analysis of the plant during a variety of potential transients to understand the primary and safety cooling system response. This study focuses on the performance of the passive safety cooling system with a dual purpose, to assess the capacity to maintain the core at safe temperatures and to assist the design process of this system to achieve this objective. The analysis requires the use of complex computational tools for simulation and verification using analytical solutions and comparisons with experimental data. This investigation builds upon previous detailed design work for the PB-AHTR components, including the core, reactivity control mechanisms and the intermediate heat exchanger, developed in 2008. In addition the study of this reference plant design employs a wealth of auxiliary information including thermal-hydraulic physical phenomena correlations for multiple geometries and thermophysical properties for the constituents of the plant. Finally, the set of performance requirements and limitations imposed from physical constrains and safety considerations provide with a criteria and metrics for acceptability of the design. The passive safety cooling system concept is turned into a detailed design as a result from this study. A methodology for the design of air-cooled passive safety systems was developed and a transient analysis of the plant, evaluating a scrammed loss of forced cooling event was performed. Furthermore, a design optimization study of the passive safety system and an approach for the validation and verification of the analysis is presented. This study demonstrates that the resulting point design responds properly to the

  18. N Reactor operational safety summary

    SciTech Connect

    Franz, G.R.; Quapp, W.J.; Ogden, D.M.

    1988-08-01

    This report is a safety summary of the N Reactor. Beginning with its conceptual design in the mid-1950`s, and throughout its 23 years of operation, continuous efforts have been made to ensure safe N Reactor operation and protection of the public health and safety. The N Reactor Updated Safety Analysis Report, completed in 1978(UNC1978), and its subsequent amendments document the safety bases of N Reactor. Following the April 1986 Chernobyl accident in the Soviet Union, a major effort to confirm N Reactor safety and further increase its safety margin was initiated. This effort, called the Safety Enhancement Program, reassessed the N Reactor using the latest accepted analysis techniques and commercial light-water reactor guidelines, where applicable. 122 refs., 38 figs., 10 tabs.

  19. Thermal reactor safety

    SciTech Connect

    Not Available

    1980-06-01

    Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport.

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

    SciTech Connect

    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. The integral design eliminates large loop piping, which significantly reduces the flow area of postulated loss of coolant accidents (LOCAs). The Westinghouse SMR containment is a high-pressure, compact design that normally operates at a partial vacuum. This facilitates heat removal from the containment during LOCA events. The containment is submerged in water which also aides the heat removal and provides an additional radionuclide filter. The Westinghouse SMR safety system design is passive, is based largely on the passive safety systems used in the AP1000{sup R} reactor, and provides mitigation of all design basis accidents without the need for AC electrical power for a period of seven days. Frequent faults, such as reactivity insertion events and loss of power events, are protected by first shutting down the nuclear reaction by inserting control rods, then providing cold, borated water through a passive, buoyancy-driven flow. Decay heat removal is provided using a layered approach that includes the passive removal of heat by the steam drum and independent passive heat removal system that transfers heat from the primary system to the environment. Less frequent faults such as loss of coolant accidents are mitigated by passive injection of a large quantity of water that is readily available inside containment. An automatic depressurization system is used to

  1. 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.

  2. Independent Safety Assessment of the TOPAZ-II space nuclear reactor power system (Revised)

    SciTech Connect

    1993-09-01

    The Independent Safety Assessment described in this study report was performed to assess the safety of the design and launch plans anticipated by the U.S. Department of Defense (DOD) in 1993 for a Russian-built, U.S.-modified, TOPAZ-II space nuclear reactor power system. Its conclusions, and the bases for them, were intended to provide guidance for the U.S. Department of Energy (DOE) management in the event that the DOD requested authorization under section 91b. of the Atomic Energy Act of 1954, as amended, for possession and use (including ground testing and launch) of a nuclear-fueled, modified TOPAZ-II. The scientists and engineers who were engaged to perform this assessment are nationally-known nuclear safety experts in various disciplines. They met with participants in the TOPAZ-II program during the spring and summer of 1993 and produced a report based on their analysis of the proposed TOPAZ-II mission. Their conclusions were confined to the potential impact on public safety and did not include budgetary, reliability, or risk-benefit analyses.

  3. Evaluation of the Safety Systems in the Next Generation Boiling Water Reactor

    NASA Astrophysics Data System (ADS)

    Cheng, Ling

    The thesis evaluates the safety systems in the next generation boiling water reactor by analyzing the main steam line break loss of coolant accident performed in the Purdue university multi-dimensional test assembly (PUMA). RELAP5 code simulations, both for the PUMA main steam line break (MSLB) case and for the simplified boiling water reactor (SBWR) MSLB case have been utilized to compare with the experiment data. The comparison shows that RELAP5 is capable to perform the safety analysis for SBWR. The comparison also validates the three-level scaling methodology applied to the design of the PUMA facility. The PUMA suppression pool mixing and condensation test data have been studied to give the detailed understanding on this important local phenomenon. A simple one dimensional integral model, which can reasonably simulate the mixing process inside suppression pool have been developed and the comparison between the model prediction and the experiment data demonstrates the model can be utilized for analyzing the suppression pool mixing process.

  4. 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.

  5. Initial Requirements for Gas-Cooled Fast Reactor (GFR) System Design, Performance, and Safety Analysis Models

    SciTech Connect

    Kevan D. Weaver; Thomas Y. C. Wei

    2004-08-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection.

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

    SciTech Connect

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

    2003-06-16

    OAK B204 The overall goal of this three-year research project was to develop a new scientific design of a compact modular 200 MWe and a full size 1200 MWe simplified boiling water reactors (SBWR). Specific objectives of this research were: (1) to perform scientific designs of the core neutronics and core thermal-hydraulics for a small capacity and full size simplified boiling water reactor, (2) to develop a passive safety system design, (3) improve and validate safety analysis code, (4) demonstrate experimentally and analytically all design functions of the safety systems for the design basis accidents (DBA) and (5) to develop the final scientific design of both SBWR systems, 200 MWe (SBWR-200) and 1200 MWe (SBWR-1200). The SBWR combines the advantages of design simplicity and completely passive safety systems. These advantages fit well within the objectives of NERI and the Department of Energy's focus on the development of Generation III and IV nuclear power. The 3-year research program was structured around seven tasks. Task 1 was to perform the preliminary thermal-hydraulic design. Task 2 was to perform the core neutronic design analysis. Task 3 was to perform a detailed scaling study and obtain corresponding PUMA conditions from an integral test. Task 4 was to perform integral tests and code evaluation for the DBA. Task 5 was to perform a safety analysis for the DBA. Task 6 was to perform a BWR stability analysis. Task 7 was to perform a final scientific design of the compact modular SBWR-200 and the full size SBWR-1200. A no cost extension for the third year was requested and the request was granted and all the project tasks were completed by April 2003. The design activities in tasks 1, 2, and 3 were completed as planned. The existing thermal-hydraulic information, core physics, and fuel lattice information was collected on the existing design of the simplified boiling water reactor. The thermal-hydraulic design were developed. Based on a detailed integral

  7. The Gas-Cooled Fast Reactor: Report on Safety System Design for Decay Heat Removal

    SciTech Connect

    K. D. Weaver; T. Marshall; T. Y. C. Wei; E. E. Feldman; M. J. Driscoll; H. Ludewig

    2003-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radiotoxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. This report addresses/discusses the decay heat removal options available to the GFR, and the current solutions. While it is possible to design a GFR with complete passive safety (i.e., reliance solely on conductive and radiative heat transfer for decay heat removal), it has been shown that the low power density results in unacceptable fuel cycle costs for the GFR. However, increasing power density results in higher decay heat rates, and the attendant temperature increase in the fuel and core. Use of active movers, or blowers/fans, is possible during accident conditions, which only requires 3% of nominal flow to remove the decay heat. Unfortunately, this requires reliance on active systems. In order to incorporate passive systems, innovative designs have been studied, and a mix of passive and active systems appears to meet the requirements for decay heat removal during accident conditions.

  8. Incorporation of Passive Safety Systems in the Generation-IV Multi-Application Small Light Water Reactor (MASLWR)

    SciTech Connect

    Modro, S. Michael; Ougouag, Abderrafi M.; Fisher, James; Weaver, Kevan

    2002-07-01

    The Idaho National Engineering and Environmental Laboratory (INEEL), Nexant Inc. and the Oregon State University (OSU) developed an innovative Multi-Application Small Light Water Reactor (MASLWR) concept. The MASLWR is a small, modular, safe, and economic natural circulation light water reactor developed with the primary goal of producing electric power, but with the flexibility to be used for water desalination or district heating with deployment in a variety of locations. The MASLWR was developed, by design, to be a safe and economic reactor concept that can be deployed in the near term by utilizing current experience and capabilities of the industry. The key features of the MASLWR concept are the extreme simplicity of the design and its passive safety systems. This paper provides an overview of safety analyses performed for the MASLWR concept and explores potential for the increase in passive safety via the implementation of new features. The results of these safety studies demonstrate that the reactor core will be provided with a stable cooling source adequate to remove decay heat without significant cladding heatup under all credible scenarios. The response of the system to accident conditions is a controlled depressurization, whereby most of the primary system blowdown occurs via the submerged ADS blowdown pathway. (authors)

  9. Providing the Basis for Innovative Improvements in Advanced LWR Reactor Passive Safety Systems Design: An Educational R&D Project

    SciTech Connect

    Brian G. Williams; Jim C. P. Liou; Hiral Kadakia; Bill Phoenix; Richard R. Schultz

    2007-02-27

    This project characterizes typical two-phase stratified flow conditions in advanced water reactor horizontal pipe sections, following activation of passive cooling systems. It provides (1) a means to educate nuclear engineering students regarding the importance of two-phase stratified flow in passive cooling systems to the safety of advanced reactor systems and (2) describes the experimental apparatus and process to measure key parameters essential to consider when designing passive emergency core cooling flow paths that may encounter this flow regime. Based on data collected, the state of analysis capabilities can be determined regarding stratified flow in advanced reactor systems and the best paths forward can be identified to ensure that the nuclear industry can properly characterize two-phase stratified flow in passive emergency core cooling systems.

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

    DOEpatents

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

    1994-05-03

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

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

    DOEpatents

    McDermott, Daniel J.; Schrader, Kenneth J.; Schulz, Terry L.

    1994-01-01

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

  12. Dynamic System Model of LS-VHTR to Estimate Design Parameter Impacts on Safety Margin and Reactor Economics

    SciTech Connect

    Qualls, A.L.; Wilson Jr, T.L.

    2006-07-01

    Early reactor analysis work for the U.S. Department of Energy's (DOE's) Liquid Salt - Very High Temperature Reactor (LS-VHTR) concept has focused primarily on detailed analyses of the core. This paper discusses ongoing analyses of the balance of plant and how it impacts overall system design. A dynamic system model of the end-to-end LS-VHTR has been developed to investigate the impact of major design parameters on systems performance, safety margin, and plant economics. The core model uses simplified thermal-hydraulic analyses to calculate four characteristic radial coolant channel parameters during transients. The core model is coupled to a multi-reheat Brayton power conversion system model through an intermediate salt-coolant loop model. A passive, safety-related heat-removal system is modeled for reactor pressure vessel protection. Critical parameters, such as peak fuel and vessel temperatures and peak temperatures and pressures in the power conversion loop, are estimated during proposed transients. The impacts of design parameters on component design requirements, safety margin, and economics are to be investigated. Transients initially analyzed will include loss-of-coolant-flow accidents. For initial transients, the axial- and radial-power profiles within the core will remain constant, with power levels decreasing in proportion to the time-dependent decay heating rate of the fuel. Later transients will represent spatial core power shifts during transients without scram. Results from simplified economic models will support relative comparisons among system design options. (authors)

  13. Probabilistic Safety Assessment of Tehran Research Reactor

    SciTech Connect

    Hosseini, Seyed Mohammad Hadi; Nematollahi, Mohammad Reza; Sepanloo, Kamran

    2004-07-01

    Probabilistic Safety Assessment (PSA) application is found to be a practical tool for research reactor safety due to intense involvement of human interactions in an experimental facility. In this paper the application of the Probabilistic Safety Assessment to the Tehran Research Reactor (TRR) is presented. The level 1 PSA application involved: Familiarization with the plant, selection of accident initiators, mitigating functions and system definitions, event tree constructions and quantification, fault tree constructions and quantification, human reliability, component failure data base development and dependent failure analysis. Each of the steps of the analysis given above is discussed with highlights from the selected results. Quantification of the constructed models is done using SAPHIRE software. This Study shows that the obtained core damage frequency for Tehran Research Reactor (8.368 E-6 per year) well meets the IAEA criterion for existing nuclear power plants (1E-4). But safety improvement suggestions are offered to decrease the most probable accidents. (authors)

  14. 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.

  15. [Reactor safety and human failure].

    PubMed

    Smidt, D

    1979-12-01

    Reactor safety is given by the reliable solution of 3 tasks: on-time shutdown, continuous decay-heat removal, safe containment. After describing the general strategy of their solution even under upset conditions the most important engineered safeguards of pressurized water reactors are summarized. The important problem of human failure is discussed in some more detail. For the example Harrisburg some difficulties, but also some technical countermeasures are illustrated. PMID:537639

  16. Thermal Reactor Safety

    SciTech Connect

    Not Available

    1980-06-01

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

  17. [Safety systems in nuclear power plants with pressurized water reactors and demonstration of their function in technical scale experiments. II].

    PubMed

    Hennies, H H

    1987-11-01

    Analyses and experiments carried out during the last decade on the sequence and consequences of accidents in German pressurized water reactors have shown that the functioning capability of the safety systems is guaranteed for the case of the MCA, the maximum credible accident. For the case of core meltdown, simulation experiments have also made it evident that the consequences remain largely restricted to the plant proper. PMID:3431585

  18. Fast Reactor Fuel Type and Reactor Safety Performance

    SciTech Connect

    R. Wigeland; J. Cahalan

    2009-09-01

    Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of ‘inherent safety’ concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and

  19. Issues affecting advanced passive light-water reactor safety analysis

    SciTech Connect

    Beelman, R.J.; Fletcher, C.D.; Modro, S.M.

    1992-08-01

    Next generation commercial reactor designs emphasize enhanced safety through improved safety system reliability and performance by means of system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs wig necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing US advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes might require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented.

  20. Issues affecting advanced passive light-water reactor safety analysis

    SciTech Connect

    Beelman, R.J.; Fletcher, C.D.; Modro, S.M.

    1992-01-01

    Next generation commercial reactor designs emphasize enhanced safety through improved safety system reliability and performance by means of system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs wig necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing US advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes might require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented.

  1. Reactor Safety Research: Semiannual report, January-June 1986: Reactor Safety Research Program

    SciTech Connect

    Not Available

    1987-05-01

    Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the technology base supporting licensing decisions.

  2. An analysis of thermionic space nuclear reactor power system: II. Merits of using safety drums for backup control

    SciTech Connect

    El-Genk, M.S.; Huimin Xue )

    1993-01-10

    An analysis is performed to investigate the merits of using the TOPAZ-II safety drums for a backup control to prevent a reactivity excursion, stabilize the reactor, and achieve steady-state power operation, following a severe hypothetical reactivity initiated accident (RIA). Such an RIA is assumed to occur during the system start-up in orbit due to a malfunction of the drive mechanism of the control drums, causing the nine drums to accidentally rotate the full 180[degree] outward. Results show that an immediate, inward rotation of the three safety drums to an angle of 80[degree] will shutdown the reactor, however, a delay time of 10 s will not only prevents a reactivity excursion, but also enables operating the reactor at a steady-state thermal power of about 33.3 kW (0.9 kW per TFE). Conversely, when the immediate rotation of the safety drums is to a larger angle of 100[degree], a steady-state operation at about 37 kW can be achieved, but a delay of 10 s causes a reactivity excursion and overheating of the TFEs. It is therefor concluded that, should the drive mechanism be modified to enable rotating the safety drums for TOPAZ-II reactor at variable speeds of and below 22.5[degree]/s, the three safety drums could be used successfully for a backup control, following an RIA. However, since the reactivity worth of the three safety drums is only $2.0, the maximum steady-state electric power achievable for the system is limited to approximately 0.25 kW, at which the fission power is about 37 kW and the emitter temperature is approximaely 1500 K. To alleviate such a limitation and enable operation at nominal design conditions (fission power of about 107 kW or a system's total electric power of 5.6 kW), the reactivity worth of the safety drums would have to be increased by at least $0.24.

  3. The next generation of power reactors - safety characteristics

    SciTech Connect

    Modro, S.M.

    1995-01-01

    The next generation of commercial nuclear power reactors is characterized by a new approach to achieving reliability of their safety systems. In contrast to current generation reactors, these designs apply passive safety features that rely on gravity-driven transfer processes or stored energy, such as gas-pressurized accumulators or electric batteries. This paper discusses the passive safety system of the AP600 and Simplified Boiling Water Reactor (SBWR) designs.

  4. Evolution of space reactor safety

    SciTech Connect

    Wetch, J.R. )

    1993-01-10

    This paper summarizes some of the major safety issues and proposed and developed solutions to real and perceived problems. Safety of the public, operating personnel, facilities and the environment has always been a driving concern of the public, government agencies, system designers and users of nuclear power and propulsion systems.

  5. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    SciTech Connect

    P. Delmolino

    2005-05-06

    The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

  6. Reactor safety for the Space Exploration Initiative

    NASA Technical Reports Server (NTRS)

    Dix, Terry E.

    1991-01-01

    A task force was created by the National Aeronautics and Space Administration to conduct a 90-day study to support efforts to determine requirements to meet the goals of the Space Exploration Initiative. The task force identified the need for a nuclear reactor to provide the electrical power required as the outpost power demands on the moon and Mars evolve into hundreds of kilowatts. A preliminary hazards analysis has been performed to examine safety aspects of nuclear reactor power systems for representative missions to the moon and Mars. Mission profiles were defined for reference lunar and Martian flights. Potential alternatives to each mission phase were also defined. Accident scenarios were qualitatively defined for the mission phases. The safety issues decay heat removal, reactor control, disposal, criticality, end-of-mission shutdown, radiation exposure, the Martian environment, high speed impact on the surfaces of the moon or Mars, and return flyby trajectories were identified.

  7. Trends in fusion reactor safety research

    NASA Astrophysics Data System (ADS)

    Herring, J. S.; Holland, D. F.; Piet, S. J.

    Fusion has the potential to be an attractive energy source. From the safety and environmental perspective, fusion must avoid concerns about catastrophic accidents and unsolvable waste disposal. In addition, fusion must achieve an acceptable level of risk from operational accidents that result in public exposure and economic loss. Finally, fusion reactors must control routine radioactive effluent, particularly tritium. Major progress in achieving this potential rests on development of low-activation materials or alternative fuels. The safety and performance of various material choices and fuels for commercial fusion reactors can be investigated relatively inexpensively through reactor design studies. These studies bring together experts in a wide range of backgrounds and force the group to either agree on a reactor design or identify areas for further study. Fusion reactors will be complex, with distributed radioactive inventories. The next generation of experiments will be critical in demonstrating that acceptable levels of safe operation can be achieved. These machines will use materials which are available today and for which a large database exists (e.g., for 316 stainless steel). Researchers have developed a good understanding of the risks associated with operation of these devices. Specifically, consequences from coolant system failures, loss of vacuum events, tritium releases, and liquid metal reactions have been studied. Recent studies go beyond next step designs and investigate commercial reactor concerns including tritium release and liquid metal reactions.

  8. Review of light water reactor safety

    SciTech Connect

    Cheng, H.S.

    1980-12-01

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

  9. 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

  10. Safety Related Investigations of the VVER-1000 Reactor Type by the Coupled Code System TRACE/PARCS

    NASA Astrophysics Data System (ADS)

    Jaeger, Wadim; Espinoza, Victor Hugo Sánchez; Lischke, Wolfgang

    This study was performed at the Institute of Reactor Safety at the Forschungszentrum Karlsruhe. It is embedded in the ongoing investigations of the international code assessment and maintenance program (CAMP) for qualification and validation of system codes like TRACE(1) and PARCS(2). The chosen reactor type used to validate these two codes was the Russian designed VVER-1000 because the OECD/NEA VVER-1000 Coolant Transient Benchmark Phase 2(3) includes detailed information of the Bulgarian nuclear power plant (NPP) Kozloduy unit 6. The post-test investigations of a coolant mixing experiment have shown that the predicted parameters (coolant temperature, pressure drop, etc.) are in good agreement with the measured data. The coolant mixing pattern, especially in the downcomer, has been also reproduced quiet well by TRACE. The coupled code system TRACE/PARCS which was applied on a postulated main steam line break (MSLB) provided good results compared to reference values and the ones of other participants of the benchmark. The results show that the developed three-dimensional nodalization of the reactor pressure vessel (RPV) is appropriate to describe the coolant mixing phenomena in the downcomer and the lower plenum of a VVER-1000 reactor. This phenomenon is a key issue for investigations of MSLB transient where the thermal hydraulics and the core neutronics are strongly linked. The simulation of the RPV and core behavior for postulated transients using the validated 3D TRACE RPV model, taking into account boundary conditions at vessel in- and outlet, indicates that the results are physically sound and in good agreement to other participant's results.

  11. 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.

  12. Small Liquid Metal Cooled Reactor Safety Study

    SciTech Connect

    Minato, A; Ueda, N; Wade, D; Greenspan, E; Brown, N

    2005-11-02

    the core life, which leads to large reactivity worth in the control systems. The conclusions from the evaluations support the high level of safety that can be achieved with small liquid metal cooled reactors using either approach.

  13. Licensed reactor nuclear safety criteria applicable to DOE reactors

    SciTech Connect

    Not Available

    1991-04-01

    The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC (Nuclear Regulatory Commission) licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor.

  14. Gas-Cooled Fast Breeder Reactor Preliminary Safety Information Document, Amendment 10. GCFR residual heat removal system criteria, design, and performance

    SciTech Connect

    Not Available

    1980-09-01

    This report presents a comprehensive set of safety design bases to support the conceptual design of the gas-cooled fast breeder reactor (GCFR) residual heat removal (RHR) systems. The report is structured to enable the Nuclear Regulatory Commission (NRC) to review and comment in the licensability of these design bases. This report also presents information concerning a specific plant design and its performance as an auxiliary part to assist the NRC in evaluating the safety design bases.

  15. Licensed reactor nuclear safety criteria applicable to DOE reactors

    SciTech Connect

    Not Available

    1993-11-01

    This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards.

  16. 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

  17. Reactor Safety Research: Semiannual report, July-December 1986

    SciTech Connect

    Not Available

    1987-11-01

    Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of the accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance and behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the tehcnology base supporting licensing decisions.

  18. Safety Systems

    ERIC Educational Resources Information Center

    Halligan, Tom

    2009-01-01

    Colleges across the country are rising to the task by implementing safety programs, response strategies, and technologies intended to create a secure environment for teachers and students. Whether it is preparing and responding to a natural disaster, health emergency, or act of violence, more schools are making campus safety a top priority. At…

  19. Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors

    SciTech Connect

    Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

    1989-10-01

    A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs.

  20. Nuclear-safety criteria and specifications for space nuclear reactors

    SciTech Connect

    Not Available

    1982-08-01

    The policy of the United States for all US nuclear power sources in space is to ensure that the probability of release of radioactive material and the amounts released are such that an undue risk is not presented, considering the benefits of the mission. The objective of this document is to provide safety criteria which a mission/reactor designer can use to help ensure that the design is acceptable from a radiological safety standpoint. These criteria encompass mission design, reactor design, and radiological impact limitation requirements for safety, and the documentation required. They do not address terrestrial operations, occupational safety or system reliability except where the systems are important for radiological safety. Specific safety specifications based on these criteria shall also be generated and made part of contractual requirements.

  1. Some views on nuclear reactor safety

    SciTech Connect

    Tanguy, P.Y.

    1995-04-01

    This document is the text of a speech given by Pierre Y. Tanguy (Electricite de France) at the 22nd Water Reactor Safety Meeting held in Bethesda, MD in 1994. He describes the EDF nuclear program in broad terms and proceeds to discuss operational safety results with EDF plants. The speaker also outlines actions to enhance safety planned for the future, and he briefly mentions French cooperation with the Chinese on the Daya Bay project.

  2. Experimental Validation of Passive Safety System Models: Application to Design and Optimization of Fluoride-Salt-Cooled, High-Temperature Reactors

    NASA Astrophysics Data System (ADS)

    Zweibaum, Nicolas

    The development of advanced nuclear reactor technology requires understanding of complex, integrated systems that exhibit novel phenomenology under normal and accident conditions. The advent of passive safety systems and enhanced modular construction methods requires the development and use of new frameworks to predict the behavior of advanced nuclear reactors, both from a safety standpoint and from an environmental impact perspective. This dissertation introduces such frameworks for scaling of integral effects tests for natural circulation in fluoride-salt-cooled, high-temperature reactors (FHRs) to validate evaluation models (EMs) for system behavior; subsequent reliability assessment of passive, natural- circulation-driven decay heat removal systems, using these validated models; evaluation of life cycle carbon dioxide emissions as a key environmental impact metric; and recommendations for further work to apply these frameworks in the development and optimization of advanced nuclear reactor designs. In this study, the developed frameworks are applied to the analysis of the Mark 1 pebble-bed FHR (Mk1 PB-FHR) under current investigation at the University of California, Berkeley (UCB). (Abstract shortened by UMI.).

  3. Analysis of Flow in Pilot Operated Safety and Relief Valve of Nuclear Reactor Coolant System

    SciTech Connect

    Kwon, Soon-Bum; Lee, Dong-Won; Kim, In-Goo; Ahn, Hyung-Joon; Kim, Hho-Jung

    2004-07-01

    When the POSRV equipped in a nuclear power plant opens in instant by a failure in coolant system of PWR, a moving shock wave generates, and propagates downstream of the valve, inducing a complicated unsteadiness. The moving shock wave may exert severe load to the structure. In this connection, a method of gradual opening of the valve is used to reduce the load acting on the wall at the downstream of the POSRV. In the present study, experiments and calculations are performed to investigate the detail unsteady flow at the various pipe units and the effect of valve opening time on the flow downstream of the valve. In calculation by using of air as working fluid, 2-dimensional, unsteady compressible Navier-Stokes equations are solved by finite volume method. It was found that when the incident shock wave passes through the pipe unit, it may experience diffraction, reflection and interaction with a vortex. Furthermore, the geometry of the pipe unit affects the reflection type of shock wave and changes the load acting on the wall of pipe unit. It was also turned out that the maximum force acting on the wall of the pipe unit becomes in order of T-junction, 108 deg. elbow and branch in magnitude, respectively. And, the results obtained that show that the rapid pressure rise due to the moving shock wave by instant POSRV valve opening is attenuated by employing the gradual opening. (authors)

  4. Reactor System Transient Code.

    Energy Science and Technology Software Center (ESTSC)

    1999-07-14

    RELAP3B describes the behavior of water-cooled nuclear reactors during postulated accidents or power transients, such as large reactivity excursions, coolant losses or pump failures. The program calculates flows, mass and energy inventories, pressures, temperatures, and steam qualities along with variables associated with reactor power, reactor heat transfer, or control systems. Its versatility allows one to describe simple hydraulic systems as well as complex reactor systems.

  5. 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.

  6. Perspectives on reactor safety. Revision 1

    SciTech Connect

    Haskin, F.E.; Camp, A.L.; Hodge, S.A.

    1997-11-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor safety concepts. The course consists of five modules: (1) the development of safety concepts; (2) severe accident perspectives; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  7. Austrian contributions to reactor safety research

    NASA Astrophysics Data System (ADS)

    Sdouz, Gert

    1992-03-01

    An overview of Austrian contributions to reactor safety research is given. Starting with licensing for the power plant Zwenkendorf and participation in the safety projects PBF (Power Burst Facility) and LOFT (Loss Of Fluid Test), the work shifted later to participation in the study of international standard problems of the OECD and the IAEA. Since 1988 the center of activity is the calculation of the source term behavior of accident sequences, especially for reactors of Russian types. As an illustration the results of a source term calculation for a TMLB' sequence are presented.

  8. Improved vortex reactor system

    DOEpatents

    Diebold, James P.; Scahill, John W.

    1995-01-01

    An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

  9. Experimental investigations of thermal-hydraulic processes arising during operation of the passive safety systems used in new projects of nuclear power plants equipped with VVER reactors

    NASA Astrophysics Data System (ADS)

    Morozov, A. V.; Remizov, O. V.; Kalyakin, D. S.

    2014-05-01

    The results obtained from experimental investigations into thermal-hydraulic processes that take place during operation of the passive safety systems used in new-generation reactor plants constructed on the basis of VVER technology are presented. The experiments were carried out on the model rigs available at the Leipunskii Institute for Physics and Power Engineering. The processes through which interaction occurs between the opposite flows of saturated steam and cold water moving in the vertical steam line of the additional system for passively flooding the core from the second-stage hydro accumulators are studied. The specific features pertinent to undeveloped boiling of liquid on a single horizontal tube heated by steam and steam-gas mixture that is typical for of the condensing operating mode of a VVER reactor steam generator are investigated.

  10. Reactor vessel support system

    DOEpatents

    Golden, Martin P.; Holley, John C.

    1982-01-01

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  11. The temperature effect and safety of the Topaz reactor

    SciTech Connect

    Artyukhov, G.Y.; Zelentsov, S.N.; Ionkin, V.I.; Kudryavtsev, V.P.; Makarenkov, Y.D.; Marin, S.N.; Pupko, V.Y.; Raskach, F.P. )

    1991-01-01

    This paper presents the results of a design calculation to compute the 3-dimensional geometry of the TOPAZ reactor. Emphasis was given to the calculation of the temperature effects and temperature reactivity coefficients. Performance of the control and safety systems, spectral and dynamic characteristics were also evaluated.(AIP)

  12. HORIZONTAL BOILING REACTOR SYSTEM

    DOEpatents

    Treshow, M.

    1958-11-18

    Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

  13. Reactor water cleanup system

    DOEpatents

    Gluntz, Douglas M.; Taft, William E.

    1994-01-01

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

  14. Reactor water cleanup system

    DOEpatents

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

    1994-12-20

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

  15. Safety significance of ATR (Advanced Test Reactor) passive safety response attributes

    SciTech Connect

    Atkinson, S.A.

    1989-01-01

    The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory was designed with some passive safety response attributes which contribute to the safety posture of the facility. The three passive safety attributes being evaluated in the paper are: (1) In-core and in-vessel natural convection cooling, (2) a passive heat sink capability of the ATR primary coolant system (PCS) for the transfer of decay power from the uninsulated piping to the confinement, and (3) gravity feed of emergency coolant makeup. The safety significance of the ATR passive safety response attributes is that the reactor can passively respond for most transients, given a reactor scram, to provide adequate decay power removal and a significant time for operator action should the normal active heat removal systems and their backup systems both fail. The ATR Interim Level 1 Probabilistic Risk Assessment (PRA) model ands results were used to evaluate the significance to ATR fuel damage frequency (or probability) of the above three passive response attributes. The results of the evaluation indicate that the first attribute is a major safety characteristic of the ATR. The second attribute has a noticeable but only minor safety significance. The third attribute has no significant influence on the ATR Level 1 PRA because of the diversity and redundancy of the ATR firewater injection system (emergency coolant system). 8 refs., 4 figs., 1 tab.

  16. Manned space flight nuclear system safety. Volume 3: Reactor system preliminary nuclear safety analysis. Part 2A: Accident model document, appendix

    NASA Technical Reports Server (NTRS)

    1972-01-01

    The detailed abort sequence trees for the reference zirconium hydride (ZrH) reactor power module that have been generated for each phase of the reference Space Base program mission are presented. The trees are graphical representations of causal sequences. Each tree begins with the phase identification and the dichotomy between success and failure. The success branch shows the mission phase objective as being achieved. The failure branch is subdivided, as conditions require, into various primary initiating abort conditions.

  17. System safety education focused on flight safety

    NASA Technical Reports Server (NTRS)

    Holt, E.

    1971-01-01

    The measures necessary for achieving higher levels of system safety are analyzed with an eye toward maintaining the combat capability of the Air Force. Several education courses were provided for personnel involved in safety management. Data include: (1) Flight Safety Officer Course, (2) Advanced Safety Program Management, (3) Fundamentals of System Safety, and (4) Quantitative Methods of Safety Analysis.

  18. Thermal state of the safety system, reactor, side reflector and shielding of the {open_quote}{open_quote}TOPAZ-2{close_quote}{close_quote} system under conditions of fire caused by a launcher accident at the launch pad

    SciTech Connect

    Grinberg, E.I.; Doschatov, V.V.; Nikolaev, V.S.; Sokolov, N.S.; Usov, V.A.

    1996-03-01

    The paper presents some results of calculational analyses performed to determine thermal state of the TOPAZ II safety system structure, radiation shielding, reactor without the side reflector, rods and inserts of the side reflector under conditions of fire at the launch pad when an accident occurs to a launch vehicle. {copyright} {ital 1996 American Institute of Physics.}

  19. Natural circulation reactor design safety analysis

    NASA Astrophysics Data System (ADS)

    Zheng, Dong

    2001-07-01

    This thesis study covers both global performance and local phenomena analyses focusing on natural circulation reactor design safety. Four important topics are included: the global SBWR design safety assessment, important local phenomena investigation, steady and transient natural circulation process study, and two-phase instability analysis. The conceptual design of the SBWR-200 is introduced in this thesis and the global performance of a natural circulation reactor is then assessed using PUMA integral test data and RELAP5 simulations. A safety assessment methodology is developed to evaluate the PUMA integral test data extrapolation and code scalability. The RELAP5 code simulation capability in low-pressure low-flow conditions is also validated. The study shows that the code is capable of predicting the global accident scenario in natural circulation reactors with reasonable accuracy, while failing to reproduce some safety related local phenomena. The natural circulation process is investigated in detail using PUMA separate effect natural circulation tests. The natural circulation flow rate and heat transfer rate have been modeled analytically and numerically. The work indicates that two-phase natural circulation has enough capability to remove decay power. However, the flow instability observed in two-phase natural circulation cases seriously challenges the feasibility of natural circulation reactor design. The instability is classified as a type of density wave instability induced by flashing. A detailed stability study is performed focusing on flashing induced instability under natural circulation condition. Various flashing phenomena have been studied and a mechanistic flashing model has been proposed and improved using a relaxation method. The developed relaxation flashing model can be applied to general two-phase non-equilibrium phenomena.

  20. Improved vortex reactor system

    DOEpatents

    Diebold, J.P.; Scahill, J.W.

    1995-05-09

    An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.

  1. Fusion reactor breeder material safety compatibility studies

    SciTech Connect

    Jeppson, D.W.; Cohen, S.; Muhlestein, L.D.

    1983-09-01

    Tritium breeder material selection for fusion reactors is strongly influenced by the desire to minimize safety and environmental concerns. Breeder material safety compatibility studies are being conducted to identify and characterize breeder-coolant-material interactions under postulated reactor accident conditions. Recently completed scoping compatibility tests indicate the following. 1. Ternary oxides (LiAlO/sub 2/, Li/sub 2/ZrO/sub 3/, Li/sub 2/SiO/sub 3/, Li/sub 4/SiO/sub 4/, and LiTiO/sub 3/) at postulated blanket operating temperatures are chemically compatible with water coolant, while liquid lithium and Li/sub 7/Pb/sub 2/ reactions with water generate heat, aerosol, and hydrogen. 2. Lithium oxide and 17Li-83Pb alloy react mildly with water requiring special precautions to control hydrogen release. 3. Liquid lithium reacts substantially, while 17Li83Pb alloy reacts mildly with concrete to produce hydrogen. 4. Liquid lithium-air reactions may present some major safety concerns. Additional scoping tests are needed, but the ternary oxides, lithium oxide, and 17Li-83Pb have definite safety advantages over liquid lithium and Li/sub 7/Pb/sub 2/. The ternary oxides present minimal safetyrelated problems when used with water as coolant, air or concrete; but they do require neutron multipliers, which may have safety compatibility concerns with surrounding materials. The combined favorable neutronics and minor safety compatibility concerns of lithium oxide and 17Li-83Pb make them prime candidates as breeder materials. Current safety efforts are directed toward assessing the compatibility of lithium oxide and the lithium-lead alloy with coolants and other materials.

  2. Reactor Safety Research Programs Quarterly Report April- June 1981

    SciTech Connect

    Edler, S. K.

    1981-09-01

    This document summarizes the work performed by Pacific Northwest laboratory (PNL} from April1 through June 30, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory {INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  3. Tornado wind-loading requirements based on risk assessment techniques (For specific reactor safety Class 1 coolant system features)

    NASA Astrophysics Data System (ADS)

    Deobald, Theodore L.; Coles, Garill A.; Smith, Gary L.

    1992-01-01

    Regulations require that nuclear power plants be protected from tornado winds. If struck by a tornado, a plant must be capable of safely shutting down and removing decay heat. Probabilistic techniques are used to show that risk to the public from the U.S. Department of Energy SP-100 reactor is acceptable without tornado hardening parts of the secondary system. Relaxed requirements for design wind loadings will result in significant cost savings. To demonstrate an acceptable level of risk, this document examines tornado-initiated accidents. The two tornado-initiated accidents examined in detail are loss of cooling resulting in core damage and loss of secondary system boundary integrity leading to sodium release. Loss of core cooling is analyzed using fault/event tree models. Loss of secondary system boundary integrity is analyzed by comparing the consequences to acceptance criteria for the release of radioactive material or alkali metal aerosol.

  4. Passive Safety Features for Small Modular Reactors

    SciTech Connect

    Ingersoll, Daniel T

    2010-01-01

    The rapid growth in the size and complexity of commercial nuclear power plants in the 1970s spawned an interest in smaller, simpler designs that are inherently or intrinsically safe through the use of passive design features. Several designs were developed, but none were ever built, although some of their passive safety features were incorporated into large commercial plant designs that are being planned or built today. In recent years, several reactor vendors are actively redeveloping small modular reactor (SMR) designs with even greater use of passive features. Several designs incorporate the ultimate in passive safety they completely eliminate specific accident initiators from the design. Other design features help to reduce the likelihood of an accident or help to mitigate the accident s consequences, should one occur. While some passive safety features are common to most SMR designs, irrespective of the coolant technology, other features are specific to water, gas, or liquid-metal cooled SMR designs. The extensive use of passive safety features in SMRs promise to make these plants highly robust, protecting both the general public and the owner/investor. Once demonstrated, these plants should allow nuclear power to be used confidently for a broader range of customers and applications than will be possible with large plants alone.

  5. Space engine safety system

    NASA Technical Reports Server (NTRS)

    Maul, William A.; Meyer, Claudia M.

    1991-01-01

    A rocket engine safety system is designed to initiate control procedures which will minimize damage to the engine and vehicle or test stand in the event of an engine failure. This report describes the features and the implementation issues associated with rocket engine safety systems. Specific concerns of safety systems applied to a space-based engine and long duration space missions are discussed. Examples of safety system features and architectures are given from recent safety monitoring investigations conducted for the Space Shuttle Main Engine and for future liquid rocket engines. Also, a general design and implementation process for rocket engine safety systems is presented.

  6. Space engine safety system

    NASA Technical Reports Server (NTRS)

    Maul, William A.; Meyer, Claudia M.

    1991-01-01

    A rocket engine safety system was designed to initiate control procedures to minimize damage to the engine or vehicle or test stand in the event of an engine failure. The features and the implementation issues associated with rocket engine safety systems are discussed, as well as the specific concerns of safety systems applied to a space-based engine and long duration space missions. Examples of safety system features and architectures are given, based on recent safety monitoring investigations conducted for the Space Shuttle Main Engine and for future liquid rocket engines. Also, the general design and implementation process for rocket engine safety systems is presented.

  7. REACTOR CONTROL SYSTEM

    DOEpatents

    MacNeill, J.H.; Estabrook, J.Y.

    1960-05-10

    A reactor control system including a continuous tape passing through a first coolant passageway, over idler rollers, back through another parallel passageway, and over motor-driven rollers is described. Discrete portions of fuel or poison are carried on two opposed active sections of the tape. Driving the tape in forward or reverse directions causes both active sections to be simultaneously inserted or withdrawn uniformly, tending to maintain a more uniform flux within the reactor. The system is particularly useful in mobile reactors, where reduced inertial resistance to control rod movement is important.

  8. Space reactor safety, 1985--1995 lessons learned

    SciTech Connect

    Marshall, A.C.

    1995-12-31

    Space reactor safety activities and decisions have evolved over the last decade. Important safety decisions have been made in the SP-100, Space Exploration Initiative, NEPSTP, SNTP, and Bimodal Space Reactor programs. In addition, international guidance on space reactor safety has been instituted. Space reactor safety decisions and practices have developed in the areas of inadvertent criticality, reentry, radiological release, orbital operation, programmatic, and policy. In general, the lessons learned point out the importance of carefully reviewing previous safety practices for appropriateness to space nuclear programs in general and to the specific mission under consideration.

  9. 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.

  10. Space Reactor Launch Safety-An Acceptably Low Risk

    NASA Astrophysics Data System (ADS)

    Weitzberg, Abraham; Wright, Steven

    2008-01-01

    Results from previous space reactor and radioisotope power source risk assessments were combined to provide a scoping assessment of the possible risks from the launch of a reactor power system for use on the surface of the moon or Mars. It is assumed that future reactor power system launches would be subject to the same rigorous safety analysis and launch approval process as past nuclear payload launches. Using the same methodology that has gained approval of past launches, it was determined that the mission risk would be 0.029 person-rem worldwide which translates to 1.5*10-5 latent health effects. It is seen that the only significant sources of radiological risks from a non-operating reactor are possible inadvertent criticality accidents and the consequences of such events have been shown to be extremely low. Passive means such as spectral shift poisons or high reactor core length/diameter ratios have been shown to be able to reduce or eliminate the possibility of the more credible criticality accidents, such as flooding or sand burial. This paper advances the premise that, for design purposes, future space reactor surface-power designs should primarily address the credible accidents and not the hypothetical accidents. For launch accidents and other safety assessments, a probabilistic risk assessment approach will have to be used to assess the safety impact of all types of accidents, including the hypothetical accidents. With this approach, the design of the system will not be burdened with design features that are based on hypothetical criticality accidents having negligible risk. Moreover, there is little chance of convincingly demonstrating that these design features can substantially reduce or eliminated the risk associated with hypothetical criticality accidents.

  11. Space Reactor Launch Safety--An Acceptably Low Risk

    SciTech Connect

    Weitzberg, Abraham; Wright, Steven

    2008-01-21

    Results from previous space reactor and radioisotope power source risk assessments were combined to provide a scoping assessment of the possible risks from the launch of a reactor power system for use on the surface of the moon or Mars. It is assumed that future reactor power system launches would be subject to the same rigorous safety analysis and launch approval process as past nuclear payload launches. Using the same methodology that has gained approval of past launches, it was determined that the mission risk would be 0.029 person-rem worldwide which translates to 1.5*10{sup -5} latent health effects. It is seen that the only significant sources of radiological risks from a non-operating reactor are possible inadvertent criticality accidents and the consequences of such events have been shown to be extremely low. Passive means such as spectral shift poisons or high reactor core length/diameter ratios have been shown to be able to reduce or eliminate the possibility of the more credible criticality accidents, such as flooding or sand burial. This paper advances the premise that, for design purposes, future space reactor surface-power designs should primarily address the credible accidents and not the hypothetical accidents. For launch accidents and other safety assessments, a probabilistic risk assessment approach will have to be used to assess the safety impact of all types of accidents, including the hypothetical accidents. With this approach, the design of the system will not be burdened with design features that are based on hypothetical criticality accidents having negligible risk. Moreover, there is little chance of convincingly demonstrating that these design features can substantially reduce or eliminated the risk associated with hypothetical criticality accidents.

  12. NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Treshow, M.

    1959-02-10

    A reactor system incorporating a reactor of the heterogeneous boiling water type is described. The reactor is comprised essentially of a core submerged adwater in the lower half of a pressure vessel and two distribution rings connected to a source of water are disposed within the pressure vessel above the reactor core, the lower distribution ring being submerged adjacent to the uppcr end of the reactor core and the other distribution ring being located adjacent to the top of the pressure vessel. A feed-water control valve, responsive to the steam demand of the load, is provided in the feedwater line to the distribution rings and regulates the amount of feed water flowing to each distribution ring, the proportion of water flowing to the submerged distribution ring being proportional to the steam demand of the load. This invention provides an automatic means exterior to the reactor to control the reactivity of the reactor over relatively long periods of time without relying upon movement of control rods or of other moving parts within the reactor structure.

  13. Safety features of subcritical fluid fueled systems

    SciTech Connect

    Bell, Charles R.

    1995-09-15

    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 qualitative 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.

  14. Safety features of subcritical fluid fueled systems

    SciTech Connect

    Bell, C.R.

    1994-09-01

    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 qualitative 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 in 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.

  15. 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.

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

    SciTech Connect

    Clarno, Kevin; Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth; Hooper, Russell Warren; Humphries, Larry LaRon

    2010-02-01

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

  17. Reactor Safety Research Programs Quarterly Report October - December 1980

    SciTech Connect

    Edler, S K

    1981-04-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from October 1 through December 31, 1980, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NOE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  18. Reactor Safety Research Programs Quarterly Report January - March 1980

    SciTech Connect

    Hagen, C. M

    1980-10-01

    This document summarizes the work performed by Pacific Northwest Laboratory from January 1 through March 31, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where serviceinduced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  19. Reactor Safety Research Programs Quarterly Report April -June 1980

    SciTech Connect

    Edler, S. K.

    1980-11-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from April 1 through June 30, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission {NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  20. Recent performance experience with US light water reactor self-actuating safety and relief valves

    SciTech Connect

    Hammer, C.G.

    1996-12-01

    Over the past several years, there have been a number of operating reactor events involving performance of primary and secondary safety and relief valves in U.S. Light Water Reactors. There are several different types of safety and relief valves installed for overpressure protection of various safety systems throughout a typical nuclear power plant. The following discussion is limited to those valves in the reactor coolant systems (RCS) and main steam systems of pressurized water reactors (PWR) and in the RCS of boiling water reactors (BWR), all of which are self-actuating having a setpoint controlled by a spring-loaded disk acting against system fluid pressure. The following discussion relates some of the significant recent experience involving operating reactor events or various testing data. Some of the more unusual and interesting operating events or test data involving some of these designs are included, in addition to some involving a number of similar events and those which have generic applicability.

  1. Moon base reactor system

    NASA Technical Reports Server (NTRS)

    Chavez, H.; Flores, J.; Nguyen, M.; Carsen, K.

    1989-01-01

    The objective of our reactor design is to supply a lunar-based research facility with 20 MW(e). The fundamental layout of this lunar-based system includes the reactor, power conversion devices, and a radiator. The additional aim of this reactor is a longevity of 12 to 15 years. The reactor is a liquid metal fast breeder that has a breeding ratio very close to 1.0. The geometry of the core is cylindrical. The metallic fuel rods are of beryllium oxide enriched with varying degrees of uranium, with a beryllium core reflector. The liquid metal coolant chosen was natural lithium. After the liquid metal coolant leaves the reactor, it goes directly into the power conversion devices. The power conversion devices are Stirling engines. The heated coolant acts as a hot reservoir to the device. It then enters the radiator to be cooled and reenters the Stirling engine acting as a cold reservoir. The engines' operating fluid is helium, a highly conductive gas. These Stirling engines are hermetically sealed. Although natural lithium produces a lower breeding ratio, it does have a larger temperature range than sodium. It is also corrosive to steel. This is why the container material must be carefully chosen. One option is to use an expensive alloy of cerbium and zirconium. The radiator must be made of a highly conductive material whose melting point temperature is not exceeded in the reactor and whose structural strength can withstand meteor showers.

  2. 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

    SciTech Connect

    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 philosophy 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.

  3. Nuclear safety criteria and specifications for space nuclear reactors

    SciTech Connect

    Not Available

    1982-08-01

    The purpose of this document is to define safety criteria which must be met to implement US safety policy for space fission reactors. These criteria provide the bases for decisions on the acceptability of specific mission and reactor design proposals. (JDH)

  4. 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)

  5. Savannah River Site K-Reactor Probabilistic Safety Assessment

    SciTech Connect

    Brandyberry, M.D.; Bailey, R.T.; Baker, W.H.; Kearnaghan, D.P.; O`Kula, K.R.; Wittman, R.S.; Woody, N.D.; Amos, C.N.; Weingardt, J.J.

    1992-12-01

    This report gives the results of a Savannah River Site (SRS) K-Reactor Probabilistic Safety Assessment (PSA). Measures of adverse consequences to health and safety resulting from representations of severe accidents in SRS reactors are presented. In addition, the report gives a summary of the methods employed to represent these accidents and to assess the resultant consequences. The report is issued to provide useful information to the U. S. Department of Energy (DOE) on the risk of operation of SRS reactors, for insights into severe accident phenomena that contribute to this risk, and in support of improved bases for other DOE programs in Heavy Water Reactor safety.

  6. Application of software to development of reactor-safety codes

    SciTech Connect

    Wilburn, N.P.; Niccoli, L.G.

    1980-09-01

    Over the past two-and-a-half decades, the application of new techniques has reduced hardware cost for digital computer systems and increased computational speed by several orders of magnitude. A corresponding cost reduction in business and scientific software development has not occurred. The same situation is seen for software developed to model the thermohydraulic behavior of nuclear systems under hypothetical accident situations. For all cases this is particularly noted when costs over the total software life cycle are considered. A solution to this dilemma for reactor safety code systems has been demonstrated by applying the software engineering techniques which have been developed over the course of the last few years in the aerospace and business communities. These techniques have been applied recently with a great deal of success in four major projects at the Hanford Engineering Development Laboratory (HEDL): 1) a rewrite of a major safety code (MELT); 2) development of a new code system (CONACS) for description of the response of LMFBR containment to hypothetical accidents, and 3) development of two new modules for reactor safety analysis.

  7. Special topics reports for the reference tandem mirror fusion breeder. Volume 2. Reactor safety assessment

    SciTech Connect

    Maya, I.; Hoot, C.G.; Wong, C.P.C.; Schultz, K.R.; Garner, J.K.; Bradbury, S.J.; Steele, W.G.; Berwald, D.H.

    1984-09-01

    The safety features of the reference fission suppressed fusion breeder reactor are presented. These include redundancy and overcapacity in primary coolant system components to minimize failure probability, an improved valve location logic to provide for failed component isolation, and double-walled coolant piping and steel guard vessel protection to further limit the extent of any leak. In addition to the primary coolant and decay heat removal system, reactor safety systems also include an independent shield cooling system, the module safety/fuel transfer coolant system, an auxiliary first wall cooling system, a psssive dump tank cooling system based on the use of heat pipes, and several lithium fire suppression systems. Safety system specifications are justified based on the results of thermal analysis, event tree construction, consequence calculations, and risk analysis. The result is a reactor design concept with an acceptably low probability of a major radioactivity release. Dose consequences of maximum credible accidents appear to be below 10CFR100 regulatory limits.

  8. Safety system status monitoring

    SciTech Connect

    Lewis, J.R.; Morgenstern, M.H.; Rideout, T.H.; Cowley, P.J.

    1984-03-01

    The Pacific Northwest Laboratory has studied the safety aspects of monitoring the preoperational status of safety systems in nuclear power plants. The goals of the study were to assess for the NRC the effectiveness of current monitoring systems and procedures, to develop near-term guidelines for reducing human errors associated with monitoring safety system status, and to recommend a regulatory position on this issue. A review of safety system status monitoring practices indicated that current systems and procedures do not adequately aid control room operators in monitoring safety system status. This is true even of some systems and procedures installed to meet existing regulatory guidelines (Regulatory Guide 1.47). In consequence, this report suggests acceptance criteria for meeting the functional requirements of an adequate system for monitoring safety system status. Also suggested are near-term guidelines that could reduce the likelihood of human errors in specific, high-priority status monitoring tasks. It is recommended that (1) Regulatory Guide 1.47 be revised to address these acceptance criteria, and (2) the revised Regulatory Guide 1.47 be applied to all plants, including those built since the issuance of the original Regulatory Guide.

  9. Flooding Experiments and Modeling for Improved Reactor Safety

    SciTech Connect

    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 line 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.

  10. The Oak Ridge Research Reactor: safety analysis: Volume 2, supplement 2

    SciTech Connect

    Hurt, S.S.

    1986-11-01

    The Oak Ridge Research Reactor Safety Analysis was last updated via ORNL-4169, Vol. 2, Supplement 1, in May of 1978. Since that date, several changes have been effected through the change-memo system described below. While these changes have involved the cooling system, the electrical system, and the reactor instrumentation and controls, they have not, for the most part, presented new or unreviewed safety questions. However, some of the changes have been based on questions or recommendations stemming from safety reviews or from reactor events at other sites. This paper discusses those changes which were judged to be safety related and which include revisions to the syphon-break system and changes related to seismic considerations which were very recently completed. The maximum hypothetical accident postulated in the original safety analysis requires dynamic containment and filtered flow for compliance with 10CFR100 limits at the site boundary.

  11. Nuclear reactor sealing system

    DOEpatents

    McEdwards, James A.

    1983-01-01

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

  12. Westinghouse independent safety review of Savannah River production reactors

    SciTech Connect

    Leggett, W.D.; McShane, W.J. ); Liparulo, N.J.; McAdoo, J.D.; Strawbridge, L.E. . Nuclear and Advanced Technology Div.); Toto, G. . Nuclear Services Div.); Fauske, H.K. ); Call, D.W. (Westinghouse Savannah R

    1989-04-01

    Westinghouse Electric Corporation has performed a safety assessment of the Savannah River production reactors (K,L, and P) as requested by the US Department of Energy. This assessment was performed between November 1, 1988, and April 1, 1989, under the transition contract for the Westinghouse Savannah River Company's preparations to succeed E.I. du Pont de Nemours Company as the US Department of Energy contractor for the Savannah River Project. The reviewers were drawn from several Westinghouse nuclear energy organizations, embody a combination of commercial and government reactor experience, and have backgrounds covering the range of technologies relevant to assessing nuclear safety. The report presents the rationale from which the overall judgment was drawn and the basis for the committee's opinion on the phased restart strategy proposed by E.I. du Pont de Nemours Company, Westinghouse, and the US Department of Energy-Savannah River. The committee concluded that it could recommend restart of one reactor at partial power upon completion of a list of recommended upgrades both to systems and their supporting analyses and after demonstration that the organization had assimilated the massive changes it will have undergone.

  13. Autonomous Flight Safety System

    NASA Technical Reports Server (NTRS)

    Ferrell, Bob; Santuro, Steve; Simpson, James; Zoerner, Roger; Bull, Barton; Lanzi, Jim

    2004-01-01

    Autonomous Flight Safety System (AFSS) is an independent flight safety system designed for small to medium sized expendable launch vehicles launching from or needing range safety protection while overlying relatively remote locations. AFSS replaces the need for a man-in-the-loop to make decisions for flight termination. AFSS could also serve as the prototype for an autonomous manned flight crew escape advisory system. AFSS utilizes onboard sensors and processors to emulate the human decision-making process using rule-based software logic and can dramatically reduce safety response time during critical launch phases. The Range Safety flight path nominal trajectory, its deviation allowances, limit zones and other flight safety rules are stored in the onboard computers. Position, velocity and attitude data obtained from onboard global positioning system (GPS) and inertial navigation system (INS) sensors are compared with these rules to determine the appropriate action to ensure that people and property are not jeopardized. The final system will be fully redundant and independent with multiple processors, sensors, and dead man switches to prevent inadvertent flight termination. AFSS is currently in Phase III which includes updated algorithms, integrated GPS/INS sensors, large scale simulation testing and initial aircraft flight testing.

  14. Nuclear reactor shutdown system

    DOEpatents

    Bhate, Suresh K.; Cooper, Martin H.; Riffe, Delmar R.; Kinney, Calvin L.

    1981-01-01

    An inherent shutdown system for a nuclear reactor having neutron absorbing rods affixed to an armature which is held in an upper position by a magnetic flux flowing through a Curie temperature material. The Curie temperature material is fixedly positioned about the exterior of an inner duct in an annular region through which reactor coolant flows. Elongated fuel rods extending from within the core upwardly toward the Curie temperature material are preferably disposed within the annular region. Upon abnormal conditions which result in high neutron flux and coolant temperature, the Curie material loses its magnetic permeability, breaking the magnetic flux path and allowing the armature and absorber rods to drop into the core, thus shutting down the fissioning reaction. The armature and absorber rods are retrieved by lowering the housing for the electromagnet forming coils which create a magnetic flux path which includes the inner duct wall. The coil housing then is raised, resetting the armature.

  15. The procedure safety system

    NASA Technical Reports Server (NTRS)

    Obrien, Maureen E.

    1990-01-01

    Telerobotic operations, whether under autonomous or teleoperated control, require a much more sophisticated safety system than that needed for most industrial applications. Industrial robots generally perform very repetitive tasks in a controlled, static environment. The safety system in that case can be as simple as shutting down the robot if a human enters the work area, or even simply building a cage around the work space. Telerobotic operations, however, will take place in a dynamic, sometimes unpredictable environment, and will involve complicated and perhaps unrehearsed manipulations. This creates a much greater potential for damage to the robot or objects in its vicinity. The Procedural Safety System (PSS) collects data from external sensors and the robot, then processes it through an expert system shell to determine whether an unsafe condition or potential unsafe condition exists. Unsafe conditions could include exceeding velocity, acceleration, torque, or joint limits, imminent collision, exceeding temperature limits, and robot or sensor component failure. If a threat to safety exists, the operator is warned. If the threat is serious enough, the robot is halted. The PSS, therefore, uses expert system technology to enhance safety thus reducing operator work load, allowing him/her to focus on performing the task at hand without the distraction of worrying about violating safety criteria.

  16. GIF sodium fast reactor project R and D on safety and operation

    SciTech Connect

    Vasile, A.; Sofu, T.; Jeong, H. Y.; Sakai, T.

    2012-07-01

    The 'Safety and Operation' project is started in 2009 within the framework of Generation-IV International Forum (GIF) Sodium Fast Reactor (SFR) research and development program. In the safety area, the project involves R and D activities on phenomenological model development and experimental programs, conceptual studies in support of the design of safety provisions, preliminary assessment of safety systems, framework and methods for analysis of safety architecture. In the operation area, the project involves R and D activities on fast reactors safety tests and analysis of reactor operations, feedback from decommissioning, in-service inspection technique development, under-sodium viewing and sodium chemistry. This paper presents a summary of such activities and the main achievements. (authors)

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

    SciTech Connect

    Forsberg, C.W.

    1990-07-01

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

  18. 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.

  19. 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)

  20. Software system safety

    NASA Technical Reports Server (NTRS)

    Uber, James G.

    1988-01-01

    Software itself is not hazardous, but since software and hardware share common interfaces there is an opportunity for software to create hazards. Further, these software systems are complex, and proven methods for the design, analysis, and measurement of software safety are not yet available. Some past software failures, future NASA software trends, software engineering methods, and tools and techniques for various software safety analyses are reviewed. Recommendations to NASA are made based on this review.

  1. Autonomous Flight Safety System

    NASA Technical Reports Server (NTRS)

    Simpson, James

    2010-01-01

    The Autonomous Flight Safety System (AFSS) is an independent self-contained subsystem mounted onboard a launch vehicle. AFSS has been developed by and is owned by the US Government. Autonomously makes flight termination/destruct decisions using configurable software-based rules implemented on redundant flight processors using data from redundant GPS/IMU navigation sensors. AFSS implements rules determined by the appropriate Range Safety officials.

  2. System Design and the Safety Basis

    SciTech Connect

    Ellingson, Darrel

    2008-05-06

    The objective of this paper is to present the Bechtel Jacobs Company, LLC (BJC) Lessons Learned for system design as it relates to safety basis documentation. BJC has had to reconcile incomplete or outdated system description information with current facility safety basis for a number of situations in recent months. This paper has relevance in multiple topical areas including documented safety analysis, decontamination & decommissioning (D&D), safety basis (SB) implementation, safety and design integration, potential inadequacy of the safety analysis (PISA), technical safety requirements (TSR), and unreviewed safety questions. BJC learned that nuclear safety compliance relies on adequate and well documented system design information. A number of PIS As and TSR violations occurred due to inadequate or erroneous system design information. As a corrective action, BJC assessed the occurrences caused by systems design-safety basis interface problems. Safety systems reviewed included the Molten Salt Reactor Experiment (MSRE) Fluorination System, K-1065 fire alarm system, and the K-25 Radiation Criticality Accident Alarm System. The conclusion was that an inadequate knowledge of system design could result in continuous non-compliance issues relating to nuclear safety. This was especially true with older facilities that lacked current as-built drawings coupled with the loss of 'historical knowledge' as personnel retired or moved on in their careers. Walkdown of systems and the updating of drawings are imperative for nuclear safety compliance. System design integration with safety basis has relevance in the Department of Energy (DOE) complex. This paper presents the BJC Lessons Learned in this area. It will be of benefit to DOE contractors that manage and operate an aging population of nuclear facilities.

  3. Programmable Electronic Safety Systems

    SciTech Connect

    Parry, R.

    1993-05-01

    Traditionally safety systems intended for protecting personnel from electrical and radiation hazards at particle accelerator laboratories have made extensive use of electromechanical relays. These systems have the advantage of high reliability and allow the designer to easily implement failsafe circuits. Relay based systems are also typically simple to design, implement, and test. As systems, such as those presently under development at the Superconducting Super Collider Laboratory (SSCL), increase in size, and the number of monitored points escalates, relay based systems become cumbersome and inadequate. The move toward Programmable Electronic Safety Systems is becoming more widespread and accepted. In developing these systems there are numerous precautions the designer must be concerned with. Designing fail-safe electronic systems with predictable failure states is difficult at best. Redundancy and self-testing are prime examples of features that should be implemented to circumvent and/or detect failures. Programmable systems also require software which is yet another point of failure and a matter of great concern. Therefore the designer must be concerned with both hardware and software failures and build in the means to assure safe operation or shutdown during failures. This paper describes features that should be considered in developing safety systems and describes a system recently installed at the Accelerator Systems String Test (ASST) facility of the SSCL.

  4. Flibe use in fusion reactors -- An initial safety assessment

    SciTech Connect

    Cadwallader, L.C.; Longhurst, G.R.

    1999-03-01

    This report is an initial effort to identify and evaluate safety issues associated with the use of Flibe (LiF-BeF{sub 2}) as a molten salt coolant for nuclear fusion power plant applications. Flibe experience in the Molten Salt Reactor Experiment is briefly reviewed. Safety issues identified include chemical toxicity, radiological issues resulting from neutron activation, and the operational concerns of handling a high temperature coolant. Beryllium compounds and fluorine pose be toxicological concerns. Some controls to protect workers are discussed. Since Flibe has been handled safely in other applications, its hazards appear to be manageable. Some safety issues that require further study are pointed out. Flibe salt interaction with strong magnetic fields should be investigated. Evolution of Flibe constituents and activation products at high temperature (i.e., will Fluorine release as a gas or remain in the molten salt) is an issue. Aerosol and tritium release from a Flibe spill requires study, as does neutronics analysis to characterize radiological doses. Tritium migration from Flibe into the cooling system is also a safety concern. Investigation of these issues will help determine the extent to which Flibe shows promise as a fusion power plant coolant or plasma-facing material.

  5. Flibe Use in Fusion Reactors - An Initial Safety Assessment

    SciTech Connect

    Cadwallader, Lee Charles; Longhurst, Glen Reed

    1999-04-01

    This report is an initial effort to identify and evaluate safety issues associated with the use of Flibe (LiF-BeF2) as a molten salt coolant for nuclear fusion power plant applications. Flibe experience in the Molten Salt Reactor Experiment is briefly reviewed. Safety issues identified include chemical toxicity, radiological issues resulting from neutron activation, and the operational concerns of handling a high temperature coolant. Beryllium compounds and fluorine pose be toxicological concerns. Some controls to protect workers are discussed. Since Flibe has been handled safely in other applications, its hazards appear to be manageable. Some safety issues that require further study are pointed out. Flibe salt interaction with strong magnetic fields should be investigated. Evolution of Flibe constituents and activation products at high temperature (i.e., will Fluorine release as a gas or remain in the molten salt) is an issue. Aerosol and tritium release from a Flibe spill requires study, as does neutronics analysis to characterize radiological doses. Tritium migration from Flibe into the cooling system is also a safety concern. Investigation of these issues will help determine the extent to which Flibe shows promise as a fusion power plant coolant or plasma-facing material.

  6. Attrition reactor system

    SciTech Connect

    Scott, C.D.; Davison, B.H.

    1993-09-28

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur. 2 figures.

  7. Attrition reactor system

    SciTech Connect

    Scott, Charles D.; Davison, Brian H.

    1993-01-01

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

  8. The Secure, Transportable, Autonomous Reactor System

    SciTech Connect

    Brown, N.W.; Hassberger, J.A.; Smith, C.; Carelli, M.; Greenspan, E.; Peddicord, K.L.; Stroh, K.; Wade, D.C.; Hill, R.N.

    1999-05-27

    The Secure, Transportable, Autonomous Reactor (STAR) system is a development architecture for implementing a small nuclear power system, specifically aimed at meeting the growing energy needs of much of the developing world. It simultaneously provides very high standards for safety, proliferation resistance, ease and economy of installation, operation, and ultimate disposition. The STAR system accomplishes these objectives through a combination of modular design, factory manufacture, long lifetime without refueling, autonomous control, and high reliability.

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

    SciTech Connect

    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.

  10. Development of electro-optical instrumentation for reactor safety studies

    SciTech Connect

    Turko, B.T.; Kolbe, W.F.; Leskovar, B.; Sun, R.K.

    1980-11-01

    The development of new electro-optical instrumentation for reactor safety studies is described. The system measures the thickness of the water film and droplet size and velocity distributions which would be encountered in the annular two-phase flow in a reactor cooling system. The water film thickness is measured by a specially designed capacitance system with a short time constant. Water droplet size and velocity are measured by a subsystem consisting of a continuously pulsed laser light source, a vidicon camera, a video recorder, and an automatic image analyzer. An endoscope system attached to the video camera is used to image the droplets. Each frame is strobed with two accurately spaced uv light pulses, from two sequentially fired nitrogen lasers. The images are stored in the video disk recorder. The modified automatic image analyzer is programmed to digitize the droplet size and velocity distributions. Many special optical, mechanical and electronic system components were designed and fabricated. They are described in detail, together with calibration charts and experimental results.

  11. 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. PMID:22021060

  12. Advanced Reactor Safety Research Division. Quarterly progress report, January 1-March 31, 1980

    SciTech Connect

    Agrawal, A.K.; Cerbone, R.J.; Sastre, C.

    1980-06-01

    The Advanced Reactor Safety Research Programs quarterly progress report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: HTGR Safety Evaluation, SSC Code Development, LMFBR Safety Experiments, and Fast Reactor Safety Code Validation.

  13. Safety status of space radioisotope and reactor power sources

    NASA Technical Reports Server (NTRS)

    Bennett, Gary L.

    1990-01-01

    The current overall safety criterion for both radioisotope and reactor power sources is containment or immobilization in the case of a reentry accident. In addition, reactors are designed to remain subcritical under conditions of land impact or water immersion. A very extensive safety test and analysis program was completed on the radioisotope thermoelectric generators (RTGs) in use on the Galileo spacecraft and planned for use on the Ulysses spacecraft. The results of this work show that the RTGs will pose little or no risk for any credible accident. The SP-100 space nuclear reactor program has begun addressing its safety criteria, and the design is planned to be such as to ensure meeting the various safety criteria. Preliminary mission risk analyses on SP-100 show the expected value population dose from postulated accidents on the reference mission to be very small. It is concluded that the current US nuclear power sources are the safest flown.

  14. 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.

  15. Generation III reactors safety requirements and the design solutions

    SciTech Connect

    Felten, P.

    2009-03-31

    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.

  16. Safety Management Systems.

    ERIC Educational Resources Information Center

    Fido, A. T.; Wood, D. O.

    This document discusses the issues that need to be considered by the education and training system as it responds to the changing needs of industry in Great Britain. Following a general introduction, the development of quality management ideas is traced. The underlying principles of safety and risk management are clarified and the implications of…

  17. Reactor vessel support system. [LMFBR

    DOEpatents

    Golden, M.P.; Holley, J.C.

    1980-05-09

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  18. Twenty-First Water Reactor Safety Information Meeting. Volume 3, Primary system integrity; Aging research, products and applications; Structural and seismic engineering; Seismology and geology: Proceedings

    SciTech Connect

    Monteleone, S.

    1994-04-01

    This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25-27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Selected papers were indexed separately for inclusion in the Energy Science and Technology Database.

  19. TOPAZ-2 Nuclear Power System safety assurance

    SciTech Connect

    Nikitin, V.P.; Ogloblin, B.G.; Lutov, Y.I.; Luppov, A.N.; Shalaev, A.I. ); Ponomarev-Stepnoi, N.N.; Usov, V.A.; Nechaev, Y.A. )

    1993-01-15

    TOPAZ-2 Nuclear Power System (NPS) safety philosophy is based on the requirement that the reactor shall not be critical during all kinds of operations prior to its start-up on the safe orbit (except for physical start-up). Potentially dangerous operation were analyzed and both computational and experimental studies were carried out.

  20. Passive safety injection system using borated water

    SciTech Connect

    Conway, Lawrence E.; Schulz, Terry L.

    1993-01-01

    A passive safety injection system relies on differences in water density to induce natural circulatory flow patterns which help maintain prescribed concentrations of boric acid in borated water, and prevents boron from accumulating in the reactor vessel and possibly preventing heat transfer.

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

    NASA Astrophysics Data System (ADS)

    Sevigny, Lawrence Michael

    to BESAFE II is discussed in Chapter 6, for example, by adding additional environmental indices such as a waste disposal index. The biggest improvement to BESAFE II would be an increase in the database of activation product mobilization for a larger spectrum of fusion reactor materials. The ultimate goal we have is for BESAFE II to become part of a systems design program which would include economic factors and allow both safety and the cost of electricity to influence design.

  2. CONVEYOR SYSTEM SAFETY ANALYSIS

    SciTech Connect

    M. Salem

    1995-06-23

    The purpose and objective of this analysis is to systematically identify and evaluate hazards related to the Yucca Mountain Project Exploratory Studies Facility (ESF) surface and subsurface conveyor system (for a list of conveyor subsystems see section 3). This process is an integral part of the systems engineering process; whereby safety is considered during planning, design, testing, and construction. A largely qualitative approach was used since a radiological System Safety Analysis is not required. The risk assessment in this analysis characterizes the accident scenarios associated with the conveyor structures/systems/components in terms of relative risk and includes recommendations for mitigating all identified risks. The priority for recommending and implementing mitigation control features is: (1) Incorporate measures to reduce risks and hazards into the structure/system/component (S/S/C) design, (2) add safety devices and capabilities to the designs that reduce risk, (3) provide devices that detect and warn personnel of hazardous conditions, and (4) develop procedures and conduct training to increase worker awareness of potential hazards, on methods to reduce exposure to hazards, and on the actions required to avoid accidents or correct hazardous conditions. The scope of this analysis is limited to the hazards related to the design of conveyor structures/systems/components (S/S/Cs) that occur during normal operation. Hazards occurring during assembly, test and maintenance or ''off normal'' operations have not been included in this analysis. Construction related work activities are specifically excluded per DOE Order 5481.1B section 4. c.

  3. FLOW SYSTEM FOR REACTOR

    DOEpatents

    Zinn, W.H.

    1963-06-11

    A reactor is designed with means for terminating the reaction when returning coolant is below a predetermined temperature. Coolant flowing from the reactor passes through a heat exchanger to a lower reservoir, and then circulates between the lower reservoir and an upper reservoir before being returned to the reactor. Means responsive to the temperature of the coolant in the return conduit terminate the chain reaction when the temperature reaches a predetermined minimum value. (AEC)

  4. Advanced reactor safety research quarterly report, October-December 1982. Volume 24

    SciTech Connect

    1984-04-01

    This report describes progress in a number of activities dealing with current safety issues relevant to both light water reactors (LWRs) and breeder reactors. The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and verification of the complex computer simulation models and codes used in accident analysis and licensing reviews. Such a program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system performance under a broad variety of abnormal conditions. Current major emphasis is focused on providing information to NRC relevant to (1) its deliberations and decisions dealing with severe LWR accidents and (2) its safety evaluation of the proposed Clinch River Breeder Reactor.

  5. Preliminary safety evaluation of the advanced burner test reactor.

    SciTech Connect

    Dunn, F. E.; Fanning, T. H.; Cahalan, J. E.; Nuclear Engineering Division

    2006-09-15

    Results of a preliminary safety evaluation of the Advanced Burner Test Reactor (ABTR) pre-conceptual design are reported. The ABTR safety design approach is described. Traditional defense-in-depth design features are supplemented with passive safety performance characteristics that include natural circulation emergency decay heat removal and reactor power reduction by inherent reactivity feedbacks in accidents. ABTR safety performance in design-basis and beyond-design-basis accident sequences is estimated based on analyses. Modeling assumptions and input data for safety analyses are presented. Analysis results for simulation of simultaneous loss of coolant pumping power and normal heat rejection are presented and discussed, both for the case with reactor scram and the case without reactor scram. The analysis results indicate that the ABTR pre-conceptual design is capable of undergoing bounding design-basis and beyond-design-basis accidents without fuel cladding failures. The first line of defense for protection of the public against release of radioactivity in accidents remains intact with significant margin. A comparison and evaluation of general safety design criteria for the ABTR conceptual design phase are presented in an appendix. A second appendix presents SASSYS-1 computer code capabilities and modeling enhancements implemented for ABTR analyses.

  6. COUNTERCURRENT FLOW LIMITATION EXPERIMENTS AND MODELING FOR IMPROVED REACTOR SAFETY

    SciTech Connect

    Vierow, Karen

    2008-09-26

    This project is investigating countercurrent flow and “flooding” phenomena in light water reactor systems to improve reactor safety of current and future reactors. To better understand the occurrence of flooding in the surge line geometry of a PWR, two experimental programs were performed. In the first, a test facility with an acrylic test section provided visual data on flooding for air-water systems in large diameter tubes. This test section also allowed for development of techniques to form an annular liquid film along the inner surface of the “surge line” and other techniques which would be difficult to verify in an opaque test section. Based on experiences in the air-water testing and the improved understanding of flooding phenomena, two series of tests were conducted in a large-diameter, stainless steel test section. Air-water test results and steam-water test results were directly compared to note the effect of condensation. Results indicate that, as for smaller diameter tubes, the flooding phenomena is predominantly driven by the hydrodynamics. Tests with the test sections inclined were attempted but the annular film was easily disrupted. A theoretical model for steam venting from inclined tubes is proposed herein and validated against air-water data. Empirical correlations were proposed for air-water and steam-water data. Methods for developing analytical models of the air-water and steam-water systems are discussed, as is the applicability of the current data to the surge line conditions. This report documents the project results from July 1, 2005 through June 30, 2008.

  7. Radiation Safety System

    SciTech Connect

    Vylet, Vaclav; Liu, James C.; Walker, Lawrence S.; /Los Alamos

    2012-04-04

    The goal of this work is to provide an overview of a Radiation safety system (RSS) designed for protection from prompt radiation hazard at accelerator facilities. RSS design parameters, functional requirements and constraints are derived from hazard analysis and risk assessment undertaken in the design phase of the facility. The two main subsystems of a RSS are access control system (ACS) and radiation control system (RCS). In this text, a common approach to risk assessment, typical components of ACS and RCS, desirable features and general design principles applied to RSS are described.

  8. Safety Assurance for Irradiating Experiments in the Advanced Test Reactor

    SciTech Connect

    T. A. Tomberlin; S. B. Grover

    2004-11-01

    The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment.

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

    SciTech Connect

    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. Major features of the newly proposed safety philosophy for the GTHTR300 are described in this article. (authors)

  10. NASA Safety Manual. Volume 3: System Safety

    NASA Technical Reports Server (NTRS)

    1970-01-01

    This Volume 3 of the NASA Safety Manual sets forth the basic elements and techniques for managing a system safety program and the technical methods recommended for use in developing a risk evaluation program that is oriented to the identification of hazards in aerospace hardware systems and the development of residual risk management information for the program manager that is based on the hazards identified. The methods and techniques described in this volume are in consonance with the requirements set forth in NHB 1700.1 (VI), Chapter 3. This volume and future volumes of the NASA Safety Manual shall not be rewritten, reprinted, or reproduced in any manner. Installation implementing procedures, if necessary, shall be inserted as page supplements in accordance with the provisions of Appendix A. No portion of this volume or future volumes of the NASA Safety Manual shall be invoked in contracts.

  11. 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.

  12. Hydrogen and water reactor safety: proceedings

    SciTech Connect

    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.

  13. Fast Reactor Spent Fuel Processing: Experience and Criticality Safety

    SciTech Connect

    Chad Pope

    2007-05-01

    This paper discusses operational and criticality safety experience associated with the Idaho National Laboratory Fuel Conditioning Facility which uses a pyrometallurgical process to treat spent fast reactor metallic fuel. The process is conducted in an inert atmosphere hot cell. The process starts with chopping metallic fuel elements into a basket. The basket is lowered into molten salt (LiCl-KCl) along with a steel mandrel. Active metal fission products, transuranic metals and sodium metal in the spent fuel undergo chemical oxidation and form chlorides. Voltage is applied between the basket, which serves as an anode, and the mandrel, which serves as a cathode, causing metallic uranium in the spent fuel to undergo electro-chemical oxidation thereby forming uranium chloride. Simultaneously at the cathode, uranium chloride undergoes electro-chemical reduction and deposits uranium metal onto the mandrel. The uranium metal and accompanying entrained salt are placed in a distillation furnace where the uranium melts forming an ingot and the entrained salt boils and subsequently condenses in a separate crucible. The uranium ingots are placed in long term storage. During the ten year operating history, over one hundred criticality safety evaluations were prepared. All criticality safety related limits and controls for the entire process are contained in a single document which required over thirty revisions to accommodate the process changes. Operational implementation of the limits and controls includes use of a near real-time computerized tracking system. The tracking system uses an Oracle database coupled with numerous software applications. The computerized tracking system includes direct fuel handler interaction with every movement of material. Improvements to this system during the ten year history include introduction of web based operator interaction, tracking of moderator materials and the development of a plethora database queries to assist in day to day

  14. NUCLEAR REACTOR FUEL SYSTEMS

    DOEpatents

    Thamer, B.J.; Bidwell, R.M.; Hammond, R.P.

    1959-09-15

    Homogeneous reactor fuel solutions are reported which provide automatic recombination of radiolytic gases and exhibit large thermal expansion characteristics, thereby providing stability at high temperatures and enabling reactor operation without the necessity of apparatus to recombine gases formed by the radiolytic dissociation of water in the fuel and without the necessity of liquid fuel handling outside the reactor vessel except for recovery processes. The fuels consist of phosphoric acid and water solutions of enriched uranium, wherein the uranium is in either the hexavalent or tetravalent state.

  15. Methods and strategies for future reactor safety goals

    NASA Astrophysics Data System (ADS)

    Arndt, Steven Andrew

    There have been significant discussions over the past few years by the United States Nuclear Regulatory Commission (NRC), the Advisory Committee on Reactor Safeguards (ACRS), and others as to the adequacy of the NRC safety goals for use with the next generation of nuclear power reactors to be built in the United States. The NRC, in its safety goals policy statement, has provided general qualitative safety goals and basic quantitative health objectives (QHOs) for nuclear reactors in the United States. Risk metrics such as core damage frequency (CDF) and large early release frequency (LERF) have been used as surrogates for the QHOs. In its review of the new plant licensing policy the ACRS has looked at the safety goals, as has the NRC. A number of issues have been raised including what the Commission had in mind when it drafted the safety goals and QHOs, how risk from multiple reactors at a site should be combined for evaluation, how the combination of a new and old reactor at the same site should be evaluated, what the criteria for evaluating new reactors should be, and whether new reactors should be required to be safer than current generation reactors. As part of the development and application of the NRC safety goal policy statement the Commissioners laid out the expectations for the safety of a nuclear power plant but did not address the risk associated with current multi-unit sites, potential modular reactor sites, and hybrid sites that could contain current generation reactors, new passive reactors, and/or modular reactors. The NRC safety goals and the QHOs refer to a "nuclear power plant," but do not discuss whether a "plant" refers to only a single unit or all of the units on a site. There has been much discussion on this issue recently due to the development of modular reactors. Additionally, the risk of multiple reactor accidents on the same site has been largely ignored in the probabilistic risk assessments (PRAs) done to date, and in most risk

  16. NUSAR: N Reactor Updated Safety Analysis Report, Amendment 21

    SciTech Connect

    Smith, G L

    1989-12-01

    The enclosed pages are Amendment 21 of the N Reactor Updated Safety Analysis Report (NUSAR). NUSAR, formerly UNI-M-90, was revised by 18 amendments that were issued by UNC Nuclear Industries, the contractor previously responsible for N Reactor operations. As of June 1987, Westinghouse Hanford Company (WHC) acquired the operations and engineering contract for N Reactor and other facilities at Hanford. The document number for NUSAR then became WHC-SP-0297. The first revision was issued by WHC as Amendment 19, prepared originally by UNC. Summaries of each of the amendments are included in NUSAR Section 1.1.

  17. Transactions of the nineteenth water reactor safety information meeting

    SciTech Connect

    Weiss, A.J.

    1991-10-01

    This report contains summaries of papers on reactor safety research to be presented at the 19th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 28--30, 1991. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, USNRC. 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 the governments and industry in Europe and Japan are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting, and are given in the order of their presentation in each session. The individual summaries have been cataloged separately.

  18. Criticality safety analysis on fissile materials in Fukushima reactor cores

    SciTech Connect

    Liu, Xudong; Lemaitre-Xavier, E.; Ahn, Joonhong; Hirano, Fumio

    2013-07-01

    The present study focuses on the criticality analysis for geological disposal of damaged fuels from Fukushima reactor cores. Starting from the basic understanding of behaviors of plutonium and uranium, a scenario sequence for criticality event is considered. Due to the different mobility of plutonium and uranium in geological formations, the criticality safety is considered in two parts: (1) near-field plutonium system and (2) far-field low enriched uranium (LEU) system. For the near-field plutonium system, a mathematical analysis for pure-solute transport was given, assuming a particular buffer material and waste form configuration. With the transport and decay of plutonium accounted, the critical mass of plutonium was compared with the initial load of a single canister. Our calculation leads us to the conclusion that our system with the initial loading being the average mass of plutonium in an assembly just before the accident is very unlikely to become critical over time. For the far-field LEU system, due to the uncertainties in the geological and geochemical conditions, calculations were made in a parametric space that covers the variation of material compositions and different geometries. Results show that the LEU system could not remain sub-critical within the entire parameter space assumed, although in the iron-rich rock, the neutron multiplicity is significantly reduced.

  19. Applicability of trends in nuclear safety analysis to space nuclear power systems

    SciTech Connect

    Bari, R.A.

    1992-10-01

    A survey is presented of some current trends in nuclear safety analysis that may be relevant to space nuclear power systems. This includes: lessons learned from operating power reactor safety and licensing; approaches to the safety design of advanced and novel reactors and facilities; the roles of risk assessment, extremely unlikely accidents, safety goals/targets; and risk-benefit analysis and communication.

  20. System code requirements for safety analysis of SBWR

    SciTech Connect

    Andersen, J.G.M.; Shiralkar, B.S.

    1994-12-31

    The simplified boiling water reactor (SBWR) being developed by General Electric Nuclear Energy is an advanced boiling water reactor relying on natural circulation during normal operation and passive safety features. The major elements of the passive safety features are the automatic depressurization of the reactor pressure vessel (RPV) following a loss-of-coolant accident (LOCA) through safety/relief valves and depressurization valves, the gravity-driven coolant system (GDCS), and the passive containment cooling system (PCCS) for residual heat removal. These passive safety systems, although based on existing technology, have generated new requirements for the computer codes used in safety and design analysis. TRACG is the computer code used for safety and design analysis for the SBWR.

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

    SciTech Connect

    Ingersoll, Daniel T

    2011-01-01

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

  2. Advanced reactor safety research. Quarterly report, April-June 1982. Volume 22

    SciTech Connect

    1983-10-01

    Overall objective of this work is to provide NRC a comprehensive data base essential to (1) defining key safety issues, (2) understanding risk-significant accident sequences, (3) developing and verifying models used in safety assessments, and (4) assuring the public that power reactor systems will not be licensed and placed in commercial service in the United States without appropriate consideration being given to their effects on health and safety. This report describes progress in a number of activities dealing with current safety issues relevant to both light water and breeder reactors. The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and verification of the complex computer simulation models and codes used in accident analysis and licensing reviews. Such a program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system performance under a broad variety of abnormal conditions. Current major emphasis is focused on providing information to NRC relevant to (1) its deliberations and decisions dealing with severe LWR accidents, and (2) its safety evaluation of the proposed Clinch River Breeder Reactor.

  3. Sodium fast reactor safety and licensing research plan. Volume II.

    SciTech Connect

    Ludewig, H.; Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A.; Phillips, J.; Zeyen, R.; Clement, B.; Garner, Frank; Walters, Leon; Wright, Steve; Ott, Larry J.; Suo-Anttila, Ahti Jorma; Denning, Richard; Ohshima, Hiroyuki; Ohno, S.; Miyhara, S.; Yacout, Abdellatif; Farmer, M.; Wade, D.; Grandy, C.; Schmidt, R.; Cahalen, J.; Olivier, Tara Jean; Budnitz, R.; Tobita, Yoshiharu; Serre, Frederic; Natesan, Ken; Carbajo, Juan J.; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Thomas, Justin; Wei, Tom; Sofu, Tanju; Flanagan, George F.; Bari, R.; Porter D.; Lambert, J.; Hayes, S.; Sackett, J.; Denman, Matthew R.

    2012-05-01

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  4. Preapplication safety evaluation report for the Power Reactor Innovative Small Module (PRISM) liquid-metal reactor. Final report

    SciTech Connect

    Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A.

    1994-02-01

    This preapplication safety evaluation report (PSER) presents the results of the preapplication desip review for die Power Reactor Innovative Small Module (PRISM) liquid-mew (sodium)-cooled reactor, Nuclear Regulatory Commission (NRC) Project No. 674. The PRISM conceptual desip was submitted by the US Department of Energy in accordance with the NRC`s ``Statement of Policy for the Regulation of Advanced Nuclear Power Plants`` (51 Federal Register 24643). This policy provides for the early Commission review and interaction with designers and licensees. The PRISM reactor desip is a small, modular, pool-type, liquid-mew (sodium)-cooled reactor. The standard plant design consists of dim identical power blocks with a total electrical output rating of 1395 MWe- Each power block comprises three reactor modules, each with a thermal rating of 471 MWt. Each module is located in its own below-grade silo and is co to its own intermediate heat transport system and steam generator system. The reactors utilize a metallic-type fuel, a ternary alloy of U-Pu-Zr. The design includes passive reactor shutdown and passive decay heat removal features. The PSER is the NRC`s preliminary evaluation of the safety features in the PRISM design, including the projected research and development programs required to support the design and the proposed testing needs. Because the NRC review was based on a conceptual design, the PSER did not result in an approval of the design. Instead it identified certain key safety issues, provided some guidance on applicable licensing criteria, assessed the adequacy of the preapplicant`s research and development programs, and concluded that no obvious impediments to licensing the PRISM design had been identified.

  5. NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Daniels, F.

    1957-10-15

    Gas-cooled solid-moderator type reactors wherein the fissionable fuel and moderator materials are each in the form of solid pebbles, or discrete particles, and are substantially homogeneously mixed in the proper proportion and placed within the core of the reactor are described. The shape of these discrete particles must be such that voids are present between them when mixed together. Helium enters the bottom of the core and passes through the voids between the fuel and moderator particles to absorb the heat generated by the chain reaction. The hot helium gas is drawn off the top of the core and may be passed through a heat exchanger to produce steam.

  6. Performance and safety parameters for the high flux isotope reactor

    SciTech Connect

    Ilas, G.; Primm III, T.

    2012-07-01

    A Monte Carlo depletion model for the High Flux Isotope Reactor (HFIR) Cycle 400 and its use in calculating parameters of relevance to the reactor performance and safety during the reactor cycle are presented in this paper. This depletion model was developed to serve as a reference for the design of a low-enriched uranium (LEU) fuel for an ongoing study to convert HFIR from high-enriched uranium (HEU) to LEU fuel; both HEU and LEU depletion models use the same methodology and ENDF/B-VII nuclear data as discussed in this paper. The calculated HFIR Cycle 400 parameters, which are compared with measurement data from critical experiments performed at HFIR, data included in the HFIR Safety Analysis Report (SAR), or data reported by previous calculations, provide a basis for verification or updating of the corresponding SAR data. (authors)

  7. Performance and Safety Parameters for the High Flux Isotope Reactor

    SciTech Connect

    Ilas, Germina; Primm, Trent

    2012-01-01

    A Monte Carlo depletion model for the High Flux Isotope Reactor (HFIR) Cycle 400 and its use in calculating parameters of relevance to the reactor performance and safety during the reactor cycle are presented in this paper. This depletion model was developed to serve as a reference for the design of a low-enriched uranium (LEU) fuel for an ongoing study to convert HFIR from high-enriched uranium (HEU) to LEU fuel; both HEU and LEU depletion models use the same methodology and ENDV/B-VII nuclear data as discussed in this paper. The calculated HFIR Cycle 400 parameters, which are compared when available with measurement data from critical experiments performed at HFIR, data included in the HFIR Safety Analysis Report (SAR), or data reported by previous calculations, provide a basis for verification or updating of the corresponding SAR data.

  8. 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.

  9. SYSTEM FOR UNLOADING REACTORS

    DOEpatents

    Rand, A.C. Jr.

    1961-05-01

    An unloading device for individual vertical fuel channels in a nuclear reactor is shown. The channels are arranged in parallel rows and underneath each is a separate supporting block on which the fuel in the channel rests. The blocks are raounted in contiguous rows on an array of parallel pairs of tracks over the bottom of the reactor. Oblong hollows in the blocks form a continuous passageway through the middle of the row of blocks on each pair of tracks. At the end of each passageway is a horizontal grappling rod with a T- or L extension at the end next to the reactor of a length to permit it to pass through the oblong passageway in one position, but when rotated ninety degrees the head will strike one of the longer sides of the oblong hollow of one of the blocks. The grappling rod is actuated by a controllable reciprocating and rotating device which extends it beyond any individual block desired, rotates it and retracts it far enough to permit the fuel in the vertical channel above the block to fall into a handling tank below the reactor.

  10. Design study on safety protection system of JSFR

    SciTech Connect

    Ishikawa, N.; Chikazawa, Y.; Fujita, K.; Yamada, Y.; Okazaki, H.; Suzuki, S.

    2012-07-01

    Development of Japan Sodium-cooled Fast Reactor (JSFR) has been progressed in Fast Reactor Cycle Technology Development (FaCT) project aiming at realizing high level of safety, reliability and economic competitiveness. For JSFR, design consideration on safety protection system has also been performed, which is essential for reactor shutdown in the case of design basis events (DBEs). In the design activity, consideration of safety protection system includes logic circuits configuration, selection of trip signals, and its setting values for reactor trip. In addition, it is necessary to evaluate the performance of the safety protection system by safety analysis taking into account the comprehensive parameter ranges. For this purpose, it has been evaluated whether adequate reactor trip signals can be ensured for satisfying safety standard regarding the fuel integrity (e.g., maximum fuel clad temperature) for DBEs. In this paper, results obtained from the design study on safety protection system of JSFR is presented focusing on the evaluation results of satisfaction of safety protection system for representative events of transient over power (TOP), loss of coolant flow (LOF) and loss of heat sink (LOHS). (authors)

  11. Evaluation of Launch Accident Safety Options for Low-Power Surface Reactors

    SciTech Connect

    Fung Poon, Cindy; Poston, David I.

    2006-01-20

    Safety options for surface reactors of less than 800 kW (thermal power) are analyzed. The concepts under consideration are heat pipe cooled reactors fueled with either uranium nitride or uranium dioxide. This study investigates the impact of launch accident criteria on the system mass, while ensuring the mechanical integrity and reliability of the system through launch accident scenarios. The four criticality scenarios analyzed for shutdown determination are dry sand surround with reflectors stripped, water submersion on concrete, water submersion with all control drums in, and the nominal shutdown reactor condition. Additionally the following two operational criteria are analyzed: reactor is warm and swelled, and reactor is warm and swelled with one drum in (where swelled includes both thermal mechanical expansion and irradiation induced swelling of the fuel)

  12. Coupled high fidelity thermal hydraulics and neutronics for reactor safety simulations

    SciTech Connect

    Vincent A. Mousseau; Hongbin Zhang; Haihua Zhao

    2008-09-01

    This work is a continuation of previous work on the importance of accuracy in the simulation of nuclear reactor safety transients. This work is qualitative in nature and future work will be more quantitative. The focus of this work will be on a simplified single phase nuclear reactor primary. The transient of interest investigates the importance of accuracy related to passive (inherent) safety systems. The transient run here will be an Unprotected Loss of Flow (ULOF) transient. Here the coolant pump is turned off and the un’SCRAM’ed reactor transitions from forced to free convection (Natural circulation). Results will be presented that show the difference that the first order in time truncation physics makes on the transient. The purpose of this document is to illuminate a possible problem in traditional reactor simulation approaches. Detailed studies need to be done on each simulation code for each transient analyzed to determine if the first order truncation physics plays an important role.

  13. Lessons learned from commercial reactor safety analysis

    NASA Astrophysics Data System (ADS)

    Fragola, J. R.

    1992-07-01

    As design concepts involving nuclear power are developed for space missions, prudence requires a consideration of the historical perspective provided by the commerical nuclear power generating station industry. This would allow the aerospace industry to take advantage of relevant historical experience, drawing from the best features and avoiding the pitfalls which appear to have stifled the growth of the commercial nuclear industry as a whole despite its comparatively admirable safety performance record. This paper provides some history of the development of commercial nuclear plant designs, and discusses the lessons which have been learned and how they apply to the space nuclear propulsion situation.

  14. A new concept of nuclear fission reactors safety

    SciTech Connect

    Petrov, Y.V.

    1993-12-31

    To develop safe nuclear energy production acceptable to the society it is proposed to use in the future strongly subcritical reactors (k=0.96-0.97) driven by proton or deuteron accelerators. The accelerator with the current of 40mA and particle energy {approximately}0.8 GeV/nucleon will provide 2 GW (th.) reactor power in fast reactor with metallic U-Pu fuel. The design, control and parameters of such a system are discussed.

  15. Preliminary safety calculations to improve the design of Molten Salt Fast Reactor

    SciTech Connect

    Brovchenko, M.; Heuer, D.; Merle-Lucotte, E.; Allibert, M.; Capellan, N.; Ghetta, V.; Laureau, A.

    2012-07-01

    Molten salt reactors are liquid fuel reactors so that they are flexible in operation but very different in the safety approach from solid fuel reactors. This study bears on the specific concept named Molten Salt Fast Reactor (MSFR). Since this new nuclear technology is in development, safety is an essential point to be considered all along the R and D studies. This paper presents the first step of the safety approach: the systematic description of the MSFR, limited here to the main systems surrounding the core. This systematic description is the basis on which we will be able to devise accidental scenarios. Thanks to the negative reactivity feedback coefficient, most accidental scenarios lead to reactor shut down. Because of the decay heat generated in the fuel salt, it must be cooled. After the description of the tools developed to calculate the residual heat, the different contributions are discussed in this study. The decay heat of fission products in the MSFR is evaluated to be low (3% of nominal power), mainly due to the reprocessing that transfers the fission products to the gas reprocessing unit. As a result, the contribution of the actinides is significant (0.5% of nominal power). The unprotected loss of heat sink transients are studied in this paper. It appears that slow transients are favorable (> 1 min) to minimize the temperature increase of the fuel salt. This work will be the basis of further safety studies as well as an essential parameter for the design of the draining system. (authors)

  16. Reactor Accident Analysis Methodology for the Advanced Test Reactor Critical Facility Documented Safety Analysis Upgrade

    SciTech Connect

    Gregg L. Sharp; R. T. McCracken

    2003-06-01

    The regulatory requirement to develop an upgraded safety basis for a DOE nuclear facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830).1 Subpart B of 10 CFR 830, “Safety Basis Requirements,” requires a contractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements.1 10 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, “Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants”2 as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology that was developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830.

  17. Reactor Accident Analysis Methodology for the Advanced Test Reactor Critical Facility Documented Safety Analysis Upgrade

    SciTech Connect

    Sharp, G.L.; McCracken, R.T.

    2003-05-13

    The regulatory requirement to develop an upgraded safety basis for a DOE Nuclear Facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830). Subpart B of 10 CFR 830, ''Safety Basis Requirements,'' requires a contractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements. 10 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, ''Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants'' as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology that was developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830.

  18. Cockpit emergency safety system

    NASA Astrophysics Data System (ADS)

    Keller, Leo

    2000-06-01

    A comprehensive safety concept is proposed for aircraft's experiencing an incident to the development of fire and smoke in the cockpit. Fire or excessive heat development caused by malfunctioning electrical appliance may produce toxic smoke, may reduce the clear vision to the instrument panel and may cause health-critical respiration conditions. Immediate reaction of the crew, safe respiration conditions and a clear undisturbed view to critical flight information data can be assumed to be the prerequisites for a safe emergency landing. The personal safety equipment of the aircraft has to be effective in supporting the crew to divert the aircraft to an alternate airport in the shortest possible amount of time. Many other elements in the cause-and-effect context of the emergence of fire, such as fire prevention, fire detection, the fire extinguishing concept, systematic redundancy, the wiring concept, the design of the power supplying system and concise emergency checklist procedures are briefly reviewed, because only a comprehensive and complete approach will avoid fatal accidents of complex aircraft in the future.

  19. Ultrasonic inspection of reactor systems

    SciTech Connect

    Majzlik, E.J. Jr.

    1989-01-01

    The subject of this presentation is ultrasonic inspection of reactor systems. This paper describes two current programs underway at Savannah River Site which provide state-of-the-art ultrasonic inspections of weld heat-affected zones in the primary cooling loop of the Savannah River Site reactors. It also describes the automated remote inspection equipment being developed and employed; briefly describe the procedures being used; and give you a general idea of the future direction of two major programs: Moderator Piping Inspection Program and the Reactor Tank Wall Weld Inspection Program. The objective of these programs is to provide inspection techniques to more fully determine the condition of the reactor primary system and provide data for prediction of maintenance needs and remaining service life. Detection and sizing of intergranular stress corrosion cracking is the focus of these programs.

  20. Traceability of Software Safety Requirements in Legacy Safety Critical Systems

    NASA Technical Reports Server (NTRS)

    Hill, Janice L.

    2007-01-01

    How can traceability of software safety requirements be created for legacy safety critical systems? Requirements in safety standards are imposed most times during contract negotiations. On the other hand, there are instances where safety standards are levied on legacy safety critical systems, some of which may be considered for reuse for new applications. Safety standards often specify that software development documentation include process-oriented and technical safety requirements, and also require that system and software safety analyses are performed supporting technical safety requirements implementation. So what can be done if the requisite documents for establishing and maintaining safety requirements traceability are not available?

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

    SciTech Connect

    Wittenbrock, N. G.

    1982-01-01

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

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

    SciTech Connect

    Diamond, D. J.; Baek, J. S.; Hanson, A. L.; Cheng, L-Y; Brown, N.; Cuadra, A.

    2015-01-30

    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 an 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.

  3. Reactor refueling containment system

    DOEpatents

    Gillett, J.E.; Meuschke, R.E.

    1995-05-02

    A method of refueling a nuclear reactor is disclosed whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced. 2 figs.

  4. Reactor refueling containment system

    DOEpatents

    Gillett, James E.; Meuschke, Robert E.

    1995-01-01

    A method of refueling a nuclear reactor whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced.

  5. Reactor core isolation cooling system

    DOEpatents

    Cooke, F.E.

    1992-12-08

    A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom. 1 figure.

  6. Reactor core isolation cooling system

    DOEpatents

    Cooke, Franklin E.

    1992-01-01

    A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom.

  7. Systems analysis of the CANDU 3 Reactor

    SciTech Connect

    Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H.

    1993-07-01

    This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ``significant to safety,`` and identification of key operator actions for the analyzed events.

  8. 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.

  9. Summary of HEDL Fusion Reactor Safety Support studies

    SciTech Connect

    Muhlestein, L.D.; Jeppson, D.W.; Barreca, J.R.

    1981-01-01

    The HEDL Fusion Reactor Safety Support studies are focused on characterizing blanket-coolant-material reactions for deuterium-tritium fusion reactor designs. The objective is to determine and examine potential safety and environmental issues associated with proposed blanket/coolant combinations under postulated accident conditions. The first studies considered liquid lithium as both blanket and coolant, and examined liquid lithium-material reactions. Liquid lithium reactions with oxygen, nitrogen, and various concretes have been characterized. Evaluations of lithium reaction extinguishment methods, lithium aerosol generation and collection, and the volatilization and transport of radioactive materials in connection with lithium-air reactions have been completed. Lithium compound blanket material reactions with water, a prime coolant candidate, have been characterized in terms of energy and gas release rates. Blanket materials considered were lithium aluminate, lithium oxide, lithium zirconate, lithium silicate, and lithium lead alloys (Li/sub 7/Pb/sub 2/ and Li/sub 17/Pb/sub 83/).

  10. Proceedings of the 1984 DOE nuclear reactor and facility safety conference. Volume II

    SciTech Connect

    Not Available

    1984-01-01

    This report is a collection of papers on reactor safety. The report takes the form of proceedings from the 1984 DOE Nuclear Reactor and Facility Safety Conference, Volume II of two. These proceedings cover Safety, Accidents, Training, Task/Job Analysis, Robotics and the Engineering Aspects of Man/Safety interfaces.

  11. Plasma reactor waste management systems

    NASA Technical Reports Server (NTRS)

    Ness, Robert O., Jr.; Rindt, John R.; Ness, Sumitra R.

    1992-01-01

    The University of North Dakota is developing a plasma reactor system for use in closed-loop processing that includes biological, materials, manufacturing, and waste processing. Direct-current, high-frequency, or microwave discharges will be used to produce plasmas for the treatment of materials. The plasma reactors offer several advantages over other systems, including low operating temperatures, low operating pressures, mechanical simplicity, and relatively safe operation. Human fecal material, sunflowers, oats, soybeans, and plastic were oxidized in a batch plasma reactor. Over 98 percent of the organic material was converted to gaseous products. The solids were then analyzed and a large amount of water and acid-soluble materials were detected. These materials could possibly be used as nutrients for biological systems.

  12. NUCLEAR REACTOR CONTROL SYSTEM

    DOEpatents

    Howard, D.F.; Motta, E.E.

    1961-06-27

    A method for controlling the excess reactivity in a nuclear reactor throughout the core life while maintaining the neutron flux distribution at the desired level is described. The control unit embodies a container having two electrodes of different surface area immersed in an electrolytic solution of a good neutron sbsorbing metal ion such as boron, gadolinium, or cadmium. Initially, the neutron absorber is plated on the larger electrode to control the greater neutron flux of a freshly refueled core. As the fuel burns up, the excess reactivity decreases and the neutron absorber is then plated onto the smaller electrode so that the number of neutrons absorbed also decreases. The excess reactivity in the core may thus be maintained without the introduction of serious perturbations in the neutron flux distributibn.

  13. Feasibility and Safety Assessment for Advanced Reactor Concepts Using Vented Fuel

    SciTech Connect

    Klein, Andrew; Matthews, Topher; Lenhof, Renae; Deason, Wesley; Harter, Jackson

    2015-01-16

    Recent interest in fast reactor technology has led to renewed analysis of past reactor concepts such as Gas Fast Reactors and Sodium Fast Reactors. In an effort to make these reactors more economic, the fuel is required to stay in the reactor for extended periods of time; the longer the fuel stays within the core, the more fertile material is converted into usable fissile material. However, as burnup of the fuel-rod increases, so does the internal pressure buildup due to gaseous fission products. In order to reach the 30 year lifetime requirements of some reactor designs, the fuel pins must have a vented-type design to allow the buildup of fission products to escape. The present work aims to progress the understanding of the feasibility and safety issues related to gas reactors that incorporate vented fuel. The work was separated into three different work-scopes: 1. Quantitatively determine fission gas release from uranium carbide in a representative helium cooled fast reactor; 2. Model the fission gas behavior, transport, and collection in a Fission Product Vent System; and, 3. Perform a safety analysis of the Fission Product Vent System. Each task relied on results from the previous task, culminating in a limited scope Probabilistic Risk Assessment (PRA) of the Fission Product Vent System. Within each task, many key parameters lack the fidelity needed for comprehensive or accurate analysis. In the process of completing each task, the data or methods that were lacking were identified and compiled in a Gap Analysis included at the end of the report.

  14. Safety of New Generation Concepts in Reference to Present Reactors

    SciTech Connect

    Rouyer, Jean-Loup; Vitton, Francis

    2004-07-01

    Present operational nuclear reactors have reached a high level of safety and reliability. Main safety principles which have allowed to obtain excellent present performances must remain the cornerstone for future projects, with adaptations on points where operating feedback shows needs for reinforcement or simplification. These basic principles are defense in depth concept and probabilistic quantification. Criteria for which specific emphasis should be put for the future are: Inertia of processes, important parameter contributing to stability, applying to neutronics as well as coolant fluids. Barriers, whose number must not be rigidly fixed, but determined by independence and margins considerations. Hazards, risk induced by external and internal hazards which should be at least, comparable to the one induced by internal accidents. On the basis of these safety criteria, the paper analyses the potentials of several reactor concepts for the future, as compared with advanced existing reactors. The six concepts selected by the Generation IV Forum (VHTR, GFR, SFR, LFR, SCWR, MSR) are appreciated. The MHTGR project which has been evaluated by the US-NRC is also considered as a reference for gas-cooled advanced concepts. (authors)

  15. Sodium fast reactor safety and licensing research plan. Volume I.

    SciTech Connect

    Sofu, Tanju; LaChance, Jeffrey L.; Bari, R.; Wigeland, Roald; Denman, Matthew R.; Flanagan, George F.

    2012-05-01

    This report proposes potential research priorities for the Department of Energy (DOE) with the intent of improving the licensability of the Sodium Fast Reactor (SFR). In support of this project, five panels were tasked with identifying potential safety-related gaps in available information, data, and models needed to support the licensing of a SFR. The areas examined were sodium technology, accident sequences and initiators, source term characterization, codes and methods, and fuels and materials. It is the intent of this report to utilize a structured and transparent process that incorporates feedback from all interested stakeholders to suggest future funding priorities for the SFR research and development. While numerous gaps were identified, two cross-cutting gaps related to knowledge preservation were agreed upon by all panels and should be addressed in the near future. The first gap is a need to re-evaluate the current procedures for removing the Applied Technology designation from old documents. The second cross-cutting gap is the need for a robust Knowledge Management and Preservation system in all SFR research areas. Closure of these and the other identified gaps will require both a reprioritization of funding within DOE as well as a re-evaluation of existing bureaucratic procedures within the DOE associated with Applied Technology and Knowledge Management.

  16. Improved Flow Modeling in Transient Reactor Safety Analysis Computer Codes

    SciTech Connect

    Holowach, M.J.; Hochreiter, L.E.; Cheung, F.B.

    2002-07-01

    A method of accounting for fluid-to-fluid shear in between calculational cells over a wide range of flow conditions envisioned in reactor safety studies has been developed such that it may be easily implemented into a computer code such as COBRA-TF for more detailed subchannel analysis. At a given nodal height in the calculational model, equivalent hydraulic diameters are determined for each specific calculational cell using either laminar or turbulent velocity profiles. The velocity profile may be determined from a separate CFD (Computational Fluid Dynamics) analysis, experimental data, or existing semi-empirical relationships. The equivalent hydraulic diameter is then applied to the wall drag force calculation so as to determine the appropriate equivalent fluid-to-fluid shear caused by the wall for each cell based on the input velocity profile. This means of assigning the shear to a specific cell is independent of the actual wetted perimeter and flow area for the calculational cell. The use of this equivalent hydraulic diameter for each cell within a calculational subchannel results in a representative velocity profile which can further increase the accuracy and detail of heat transfer and fluid flow modeling within the subchannel when utilizing a thermal hydraulics systems analysis computer code such as COBRA-TF. Utilizing COBRA-TF with the flow modeling enhancement results in increased accuracy for a coarse-mesh model without the significantly greater computational and time requirements of a full-scale 3D (three-dimensional) transient CFD calculation. (authors)

  17. Microprocessor tester for the treat upgrade reactor trip system

    SciTech Connect

    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 system is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations.

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

    SciTech Connect

    Vilim, R. B.; Park, Y. S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J.

    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 response 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.

  19. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    SciTech Connect

    Ryskamp, J.M.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG&G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

  20. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    SciTech Connect

    Ryskamp, J.M.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

  1. Solvent refined coal reactor quench system

    DOEpatents

    Thorogood, Robert M.

    1983-01-01

    There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream.

  2. Solvent refined coal reactor quench system

    DOEpatents

    Thorogood, R.M.

    1983-11-08

    There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream. 1 fig.

  3. Rapid starting methanol reactor system

    DOEpatents

    Chludzinski, Paul J.; Dantowitz, Philip; McElroy, James F.

    1984-01-01

    The invention relates to a methanol-to-hydrogen cracking reactor for use with a fuel cell vehicular power plant. The system is particularly designed for rapid start-up of the catalytic methanol cracking reactor after an extended shut-down period, i.e., after the vehicular fuel cell power plant has been inoperative overnight. Rapid system start-up is accomplished by a combination of direct and indirect heating of the cracking catalyst. Initially, liquid methanol is burned with a stoichiometric or slightly lean air mixture in the combustion chamber of the reactor assembly. The hot combustion gas travels down a flue gas chamber in heat exchange relationship with the catalytic cracking chamber transferring heat across the catalyst chamber wall to heat the catalyst indirectly. The combustion gas is then diverted back through the catalyst bed to heat the catalyst pellets directly. When the cracking reactor temperature reaches operating temperature, methanol combustion is stopped and a hot gas valve is switched to route the flue gas overboard, with methanol being fed directly to the catalytic cracking reactor. Thereafter, the burner operates on excess hydrogen from the fuel cells.

  4. 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.

  5. System safety engineering analysis handbook

    NASA Technical Reports Server (NTRS)

    Ijams, T. E.

    1972-01-01

    The basic requirements and guidelines for the preparation of System Safety Engineering Analysis are presented. The philosophy of System Safety and the various analytic methods available to the engineering profession are discussed. A text-book description of each of the methods is included.

  6. Monte Carlo verification of point kinetics for safety analysis of nuclear reactors

    SciTech Connect

    Valentine, T.E.; Mihalczo, J.T.

    1995-06-01

    Monte Carlo neutron transport methods can be used to verify the applicability of point kinetics for safety analysis of nuclear reactors. KENO-NR was used to obtain the transfer function of the Advanced Neutron Source reactor and the time delay between the core power production and the external detectors, a parameter of interest to the safety systems design. The good agreement between the Monte Carlo generated transfer function and the point kinetics transfer function validates that the uncommon ANS geometry does not preclude the use of point kinetics in the frequency range that was investigated. Various features of the power spectral densities also demonstrated the applicability of point kinetics. The time delay was obtained from the cross-power spectral density (CPSD) and is {approximately}15 ms. These analyses show that frequency analysis can be used experimentally to investigate the validity of the use of point kinetics models in critical experiments or zero power testing of reactors.

  7. Evaluating software for safety systems in nuclear power plants

    SciTech Connect

    Lawrence, J.D.; Persons, W.L.; Preckshot, G.G.; Gallagher, J.

    1994-01-11

    In 1991, LLNL was asked by the NRC to provide technical assistance in various aspects of computer technology that apply to computer-based reactor protection systems. This has involved the review of safety aspects of new reactor designs and the provision of technical advice on the use of computer technology in systems important to reactor safety. The latter includes determining and documenting state-of-the-art subjects that require regulatory involvement by the NRC because of their importance in the development and implementation of digital computer safety systems. These subjects include data communications, formal methods, testing, software hazards analysis, verification and validation, computer security, performance, software complexity and others. One topic software reliability and safety is the subject of this paper.

  8. Progress toward international agreement to improve reactor safety

    SciTech Connect

    Lieberman, J.I.; Graham, B.

    1993-05-14

    Representatives of nearly one-half of the 114 member states of the International Atomic Energy Agency (IAEA), including the United States, have participated in the development of an international nuclear safety conventions proposed multilateral treaty to improve civil nuclear power reactor safety. A preliminary draft of the convention has been developed (referred to as the draft convention for this report), but discussions are continuing, and when the final convention text will be completed and presented to IAEA member states for signature is uncertain. This report responds to the former and current Chairman`s request that we provide information on the development of the nuclear safety convention, including a discussion of (1) the draft convention`s scope and objectives, (2) how the convention will be implemented and monitored, (3) the views of selected country representatives on what provisions should be included in the draft convention, and (4) the convention`s potential benefits and limitations.

  9. Reactor Safety Research Programs Quarterly Report July - September 1981

    SciTech Connect

    Edler, S. K.

    1982-01-01

    This document summarizes the work performed by Pacific Northwest laboratory (PNL) from July 1 through September 30, 1981, for the Division of Accident Evaluation, U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR} steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL), Idaho Falls, Idaho. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  10. Reactor Safety Research Programs Quarterly Report October - December 1981

    SciTech Connect

    Edler, S. K.

    1982-03-01

    This document summarizes the work performed by Pacific Northwest laboratory (PNL) from October 1 through December 31, 1981, for the Division of Accident Evaluation, U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where serviceinduced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and post accident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL), Idaho Falls, Idaho. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  11. 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.

  12. 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.

  13. CANDU reactors, their regulation in Canada, and the identification of relevant NRC safety issues

    SciTech Connect

    Charak, I.; Kier, P.H.

    1995-04-01

    Atomic Energy of Canada, Limited (AECL) and its subsidiary in the US, are considering submitting the CANDU 3 design for standard design certification under 10 CFR Part 52. CANDU reactors are pressurized heavy water power reactors. They have some substantially different safety responses and safety systems than the LWRs that the commercial power reactor licensing regulations of the US Nuclear Regulatory Commission (NRC) have been developed to deal with. In this report, the authors discuss the basic design characteristics of CANDU reactors, specifically of the CANDU 3 where possible, and some safety-related consequences of these characteristics. The authors also discuss the Canadian regulatory provisions, and the CANDU safety systems that have evolved to satisfy the Canadian regulatory requirements as of December 1992. Finally, the authors identify NRC regulations, mainly in 10 CFR Parts 50 and 100, with issues for CANDU 3 reactor designs. In all, eleven such regulatory issues are identified. They are: (1) the ATWS rule ({section}50.62); (2) station blackout ({section}50.63); (3) conformance with Standard Review Plan (SRP); (4) appropriateness of the source term ({section}50.34(f) and {section}100.11); (5) applicability of reactor coolant pressure boundary (RCPB) requirements ({section}50.55a, etc); (6) ECCS acceptance criteria ({section}50.46)(b); (7) combustible gas control ({section}50.44, etc); (8) power coefficient of reactivity (GDC 11); (9) seismic design (Part 100); (10) environmental impacts of the fuel cycle ({section}51.51); and (11) (standards {section}50.55a).

  14. Adjoint-based sensitivity analysis for reactor safety applications

    SciTech Connect

    Parks, C.V.

    1986-08-01

    The application and usefulness of an adjoint-based methodology for performing sensitivity analysis on reactor safety computer codes is investigated. The adjoint-based methodology, referred to as differential sensitivity theory (DST), provides first-order derivatives of the calculated quantities of interest (responses) with respect to the input parameters. The basic theoretical development of DST is presented along with the needed general extensions for consideration of model discontinuities and a variety of useful response definitions. A simple analytic problem is used to highlight the general DST procedures. finally, DST procedures presented in this work are applied to two highly nonlinear reactor accident analysis codes: (1) FASTGAS, a relatively small code for analysis of a loss-of-decay-heat-removal accident in a gas-cooled fast reactor, and (2) an existing code called VENUS-II which has been employed for analyzing the core disassembly phase of a hypothetical fast reactor accident. The two codes are different both in terms of complexity and in terms of the facets of DST which can be illustrated. Sensitivity results from the adjoint codes ADJGAS and VENUS-ADJ are verified with direct recalcualtions using perturbed input parameters. The effectiveness of the DST results for parameter ranking, prediction of response changes, and uncertainty analysis are illustrated. The conclusion drawn from this study is that DST is a viable, cost-effective methodology for accurate sensitivity analysis. In addition, a useful sensitivity tool for use in the fast reactor safety area has been developed in VENUS-ADJ. Future work needs to concentrate on combining the accurate first-order derivatives/results from DST with existing methods (based solely on direct recalculations) for higher-order response surfaces.

  15. Small space reactor power systems for unmanned solar system exploration missions

    SciTech Connect

    Bloomfield, H.S.

    1987-12-01

    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.

  16. 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.

  17. Regulatory analysis for the resolution of Generic Safety Issue 105: Interfacing system loss-of-coolant accident in light-water reactors

    SciTech Connect

    Not Available

    1993-07-01

    An interfacing systems loss of coolant accident (ISLOCA) involves failure or improper operation of pressure isolation valves (PIVs) that compose the boundary between the reactor coolant system and low-pressure rated systems. Some ISLOCAs can bypass containment and result in direct release of fission products to the environment. A cost/benefit evaluation, using three PWR analyses, calculated the benefit of two potential modifications to the plants. Alternative 1 is improved plant operations to optimize the operator`s performance and reduce human error probabilities. Alternative 2 adds pressure sensing devices, cabling, and instrumentation between two PIVs to provide operators with continuous monitoring of the first PIV. These two alternatives were evaluated for the base case plants (Case 1) and for each plant, assuming the plants had a particular auxiliary building design in which severe flooding would be a problem if an ISLOCA occurred. The auxiliary building design (Case 2) was selected from a survey that revealed a number of designs with features that provided less than optimal resistance to ECCS equipment loss caused by a ISLOCA-induced environment. The results were judged not to provide sufficient basis for generic requirements. It was concluded that the most viable course of action to resolve Generic Issue 105 is licensee participation in individual plant examinations (IPEs).

  18. 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.

  19. Design of a 25-kWe Surface Reactor System Based on SNAP Reactor Technologies

    SciTech Connect

    Dixon, David D.; Hiatt, Matthew T.; Poston, David I.; Kapernick, Richard J.

    2006-01-20

    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.

  20. Tritium management in a fusion reactor--safety, handling and economical issues--

    SciTech Connect

    Tanabe, Tetsuo

    2009-02-19

    In order to establish a D-T fusion reactor as an energy source, it is not enough to have a DT burning plasma, and economical conversion of fusion energy to electricity and/or heat, a large enough margin of tritium breeding and tritium safety must be simultaneously achieved. In particular, handling of huge amount of tritium needs significant efforts to ensure that the radiation dose of radiological workers and of the public is below the limits specified by the International Commission on Radiological Protection (ICRP). In this paper, after the introduction of tritium as a fuel of DT reactors and as a radioisotope of hydrogen, tritium safety issues in fuel cycle and blanket systems are summarized. In particular, in-vessel tritium inventory, the most important and uncertain tritium safety issue, is discussed in detail.

  1. Dynamical Safety Analysis of the SABR Fusion-Fission Hybrid Reactor

    NASA Astrophysics Data System (ADS)

    Sumner, Tyler; Stacey, Weston; Ghiaassian, Seyed

    2009-11-01

    A hybrid fusion-fission reactor for the transmutation of spent nuclear fuel is being developed at Georgia Tech. The Subcritical Advanced Burner Reactor (SABR) is a 3000 MWth sodium-cooled, metal TRU-Zr fueled fast reactor driven by a tokamak fusion neutron source based on ITER physics and technology. We are investigating the accident dynamics of SABR's coupled fission, fusion and heat removal systems to explore the safety characteristics of a hybrid reactor. Possible accident scenarios such as loss of coolant mass flow (LOFA), of power (LOPA) and of heat sink (LOHSA), as well as inadvertent reactivity insertions and fusion source excursion are being analyzed using the RELAP5-3D code, the ATHENA version of which includes liquid metal coolants.

  2. Physics of reactor safety. Quarterly report, October-December 1980. Volume IV

    SciTech Connect

    Not Available

    1981-02-01

    The work in the Applied Physics Division includes reports on reactor safety modeling and assessment by members of the Reactor Safety Appraisals Section. Work on reactor core thermal-hydraulics is performed in ANL's Components Technology Division, emphasizing 3-dimensional code development for LMFBR accidents under natural convection conditions.

  3. Recent developments in Topaz II reactor safety assessments

    SciTech Connect

    Marshall, A.C.

    1993-07-01

    In December 1991, the Strategic Defense Initiative Organization (SDIO) decided to investigate the possibility of a US launch of a Russian Topaz II space nuclear power system. The primary mission goal would be to demonstrate and evaluate Nuclear Electric Propulsion technology to establish a capability for future civilian and military missions. A preliminary nuclear safety assessment, involving selected safety analyses, was initiated to determine whether or not a space mission could be conducted safely and within budget constraints. This paper describes the preliminary safety assessment results and the nuclear safety program now being established for the Nuclear Electric Propulsion Space Test Program (NEPSTP).

  4. Determination of core design thermal safety limits for a two-loop pressurized water reactor

    SciTech Connect

    Kostadinov, V.

    1996-04-01

    Results are given of independent research of core thermal design limits for the Nuklearna Elektrarna Krsko (NEK) nuclear power plant; procedures for two-loop pressurized water reactor plant core design safety limit calculation are used. Emphasis is placed on researching the vessel exit boiling and the hot-channel exit quality limits and their impact on the maximum available design safety operating range and thermal operating margin of the NEK reactor core. For this purpose, the LIMITS computer code is developed. Based on the modified, well-tried COBRA-IV-I computer code, the departure of nuclear boiling ratio core safety limits are calculated. The original results complement well those of the NEK Final Safety Analysis Report. The procedures and the methods for determining the reactor core design thermal limits are successfully proven despite the unavailability of proprietary data, different models, and computer codes. In addition to the acquired capability of in-house independent checking of the vendor`s results, the bases are set for further independent analyses of the limiting safety system settings for the NEK core.

  5. Fusion-reactor blanket-material safety-compatibility studies

    SciTech Connect

    Jeppson, D.W.; Muhlestein, L.D.; Keough, R.F.; Cohen, S.

    1982-11-01

    Blanket material selection for fusion reactors is strongly influenced by the desire to minimize safety and environmental concerns. Blanket material safety compatibility studies are being conducted to identify and characterize blanket-coolant-material interactions under postulated reactor accident conditions. Recently completed scoping compatibility tests indicate that : (1) ternary oxides (LiAlO/sub 2/, Li/sub 2/ZrO/sub 3/, Li/sub 2/SiO/sub 3/, Li/sub 4/SiO/sub 4/ and LiTiO/sub 3/) at postulated blanket operating temperatures are compatible with water coolant, while liquid lithium and Li/sub 7/Pb/sub 2/ alloy reactions with water generate heat, aerosol and hydrogen; (2) lithium oxide and Li/sub 17/Pb/sub 83/ alloy react mildly with water requiring special precautions to control hydrogen release; (3) liquid lithium reacts substantially, while Li/sub 17/Pb/sub 83/ alloy reacts mildly with concrete to produce hydrogen; and (4) liquid lithium-air reactions present some major safety concerns.

  6. Software Safety Risk in Legacy Safety-Critical Computer Systems

    NASA Technical Reports Server (NTRS)

    Hill, Janice L.; Baggs, Rhoda

    2007-01-01

    Safety Standards contain technical and process-oriented safety requirements. Technical requirements are those such as "must work" and "must not work" functions in the system. Process-Oriented requirements are software engineering and safety management process requirements. Address the system perspective and some cover just software in the system > NASA-STD-8719.13B Software Safety Standard is the current standard of interest. NASA programs/projects will have their own set of safety requirements derived from the standard. Safety Cases: a) Documented demonstration that a system complies with the specified safety requirements. b) Evidence is gathered on the integrity of the system and put forward as an argued case. [Gardener (ed.)] c) Problems occur when trying to meet safety standards, and thus make retrospective safety cases, in legacy safety-critical computer systems.

  7. Adjoint-based sensitivity analysis for reactor-safety applications

    SciTech Connect

    Parks, C.V.

    1985-01-01

    The application and usefulness of an adjoint-based methodology for performing sensitivity analysis on reactor safety computer codes is investigated. The adjoint-based methodology, referred to as differential sensitivity theory (DST), provides first-order derivatives of the calculated quantities of interest (responses) with respect to the input parameters. The basic theoretical development of DST is presented along with the needed general extensions for consideration of model discontinuities and a variety of useful response definitions. A simple analytic problem is used to highlight the general DST procedures. Finally, DST procedures presented in this work are applied to two highly nonlinear reactor accident analysis codes: (1) FASTGAS, a relatively small code for analysis of loss-of-decay-heat-removal accident in a gas-cooled fast reactor, and (2) an existing code called VENUS-II which is typically employed for analyzing the core disassembly phase of a hypothetical fast reactor accident. The two codes are different both in terms of complexity and in terms of the facets of DST which can be illustrated. Sensitivity results from the adjoint codes ADJGAS and VENUS-ADJ are verified with direct recalculations using perturbed input parameters. The effectiveness of the DST results for parameter ranking, prediction of response changes, and uncertainty analysis are illustrated. The conclusion drawn from this study is that DST is a viable, cost-effective methodology for accurate sensitivity analysis.

  8. NASA aviation safety reporting system

    NASA Technical Reports Server (NTRS)

    1979-01-01

    The human factors frequency considered a cause of or contributor to hazardous events onboard air carriers are examined with emphasis on distractions. Safety reports that have been analyzed, processed, and entered into the aviation safety reporting system data base are discussed. A sampling of alert bulletins and responses to them is also presented.

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

    SciTech Connect

    David Petti; Jim Kinsey; Dave Alberstein

    2014-01-01

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

  10. Integrated systems analysis of the PIUS reactor

    SciTech Connect

    Fullwood, F.; Kroeger, P.; Higgins, J.

    1993-11-01

    Results are presented of a systems failure analysis of the PIUS plant systems that are used during normal reactor operation and postulated accidents. This study was performed to provide the NRC with an understanding of the behavior of the plant. The study applied two diverse failure identification methods, Failure Modes Effects & Criticality Analysis (FMECA) and Hazards & Operability (HAZOP) to the plant systems, supported by several deterministic analyses. Conventional PRA methods were also used along with a scheme for classifying events by initiator frequency and combinations of failures. Principal results of this study are: (a) an extensive listing of potential event sequences, grouped in categories that can be used by the NRC, (b) identification of support systems that are important to safety, and (c) identification of key operator actions.

  11. Proceedings of the Twenty-First Water Reactor Safety Information Meeting: Volume 1, Plenary session; Advanced reactor research; advanced control system technology; advanced instrumentation and control hardware; human factors research; probabilistic risk assessment topics; thermal hydraulics; thermal hydraulic research for advanced passive LWRs

    SciTech Connect

    Monteleone, S.

    1994-04-01

    This three-volume report contains 90 papers out of the 102 that were presented at the Twenty-First Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 25--27, 1993. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Germany, Japan, Russia, Switzerland, Taiwan, and United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. Individual papers have been cataloged separately. This document, Volume 1 covers the following topics: Advanced Reactor Research; Advanced Instrumentation and Control Hardware; Advanced Control System Technology; Human Factors Research; Probabilistic Risk Assessment Topics; Thermal Hydraulics; and Thermal Hydraulic Research for Advanced Passive Light Water Reactors.

  12. System safety education focused on system management

    NASA Technical Reports Server (NTRS)

    Grose, V. L.

    1971-01-01

    System safety is defined and characteristics of the system are outlined. Some of the principle characteristics include role of humans in hazard analysis, clear language for input and output, system interdependence, self containment, and parallel analysis of elements.

  13. Generic safety insights for inspection of boiling water reactors

    SciTech Connect

    Higgins, J.C.; Taylor, J.H.; Fresco, A.N.; Hillman, B.M.

    1987-01-01

    As the number of operating nuclear power plants (NPPs) increases, safety inspection has increased in importance. Over the last 2 yr, probabilistic risk assessment (PRA) techniques have been developed to aid in the inspection process. Broad interest in generic PRA-based methods has arisen in the past year, since only approx. 25% of the US nuclear power plants have completed PRAs, and also, inspectors want PRA-based tools for these plants. This paper describes the Brookhaven National Lab. program to develop generic boiling water reactor (BWR) PRA-based inspection insights or inspection guidance designed to be applied to plants without PRAs.

  14. Study of Cost Effective Large Advanced Pressurized Water Reactors that Employ Passive Safety Features

    SciTech Connect

    Winters, J. W.; Corletti, M. M.; Hayashi, Y.

    2003-11-12

    A report of DOE sponsored portions of AP1000 Design Certification effort. On December 16, 1999, The United States Nuclear Regulatory Commission issued Design Certification of the AP600 standard nuclear reactor design. This culminated an 8-year review of the AP600 design, safety analysis and probabilistic risk assessment. The AP600 is a 600 MWe reactor that utilizes passive safety features that, once actuated, depend only on natural forces such as gravity and natural circulation to perform all required safety functions. These passive safety systems result in increased plant safety and have also significantly simplified plant systems and equipment, resulting in simplified plant operation and maintenance. The AP600 meets NRC deterministic safety criteria and probabilistic risk criteria with large margins. A summary comparison of key passive safety system design features is provided in Table 1. These key features are discussed due to their importance in affecting the key thermal-hydraulic phenomenon exhibited by the passive safety systems in critical areas. The scope of some of the design changes to the AP600 is described. These changes are the ones that are important in evaluating the passive plant design features embodied in the certified AP600 standard plant design. These design changes are incorporated into the AP1000 standard plant design that Westinghouse is certifying under 10 CFR Part 52. In conclusion, this report describes the results of the representative design certification activities that were partially supported by the Nuclear Energy Research Initiative. These activities are unique to AP1000, but are representative of research activities that must be driven to conclusion to realize successful licensing of the next generation of nuclear power plants in the United States.

  15. Reactor vessel annealing system

    DOEpatents

    Miller, Phillip E.; Katz, Leonoard R.; Nath, Raymond J.; Blaushild, Ronald M.; Tatch, Michael D.; Kordalski, Frank J.; Wykstra, Donald T.; Kavalkovich, William M.

    1991-01-01

    A system for annealing a vessel (14) in situ by heating the vessel (14) to a defined temperature, composed of: an electrically operated heater assembly (10) insertable into the vessel (14) for heating the vessel (14) to the defined temperature; temperature monitoring components positioned relative to the heater assembly (10) for monitoring the temperature of the vessel (14); a controllable electric power supply unit (32-60) for supplying electric power required by the heater assembly (10); a control unit (80-86) for controlling the power supplied by the power supply unit (32-60); a first vehicle (2) containing the power supply unit (32-60); a second vehicle (4) containing the control unit (80-86); power conductors (18,22) connectable between the power supply unit (32-60) and the heater unit (10) for delivering the power supplied by the power supply unit (32-60) to the heater assembly (10); signal conductors (20,24) connectable between the temperature monitoring components and the control unit (80-86) for delivering temperature indicating signals from the temperature monitoring components to the control unit (80-86); and control conductors (8) connectable between the control unit (80-86) and the power supply unit (32-60) for delivering to the power supply unit (32-60) control signals for controlling the level of power supplied by the power supply unit (32-60) to the heater assembly (10).

  16. New generation of NPP with boiling water reactor of improved safety

    SciTech Connect

    Adamov, E.O.; Kuklin, A.N.; Mityaev, Yu.I.; Mikhan, V.I.; Tokarev, Yu.I.; Cherkashov, Yu.M.; Sokolov, I.N.; Iljin, Yu.V.; Pakh, E.E.; Abramov, V.I.

    1993-12-31

    The nuclear power plants with boiling water reactors of improved safety are being developed. There is 26 years of operating experience with the plant VK-50 in Dimitrovgrad. The design and operation of the BWR reactors are described.

  17. Safety Aspects of Big Cryogenic Systems Design

    NASA Astrophysics Data System (ADS)

    Chorowski, M.; Fydrych, J.; Poliński, J.

    2010-04-01

    Superconductivity and helium cryogenics are key technologies in the construction of large scientific instruments, like accelerators, fusion reactors or free electron lasers. Such cryogenic systems may contain more than hundred tons of helium, mostly in cold and high-density phases. In spite of the high reliability of the systems, accidental loss of the insulation vacuum, pipe rupture or rapid energy dissipation in the cold helium can not be overlooked. To avoid the danger of over-design pressure rise in the cryostats, they need to be equipped with a helium relief system. Such a system is comprised of safety valves, bursting disks and optionally cold or warm quench lines, collectors and storage tanks. Proper design of the helium safety relief system requires a good understanding of worst case scenarios. Such scenarios will be discussed, taking into account different possible failures of the cryogenic system. In any case it is necessary to estimate heat transfer through degraded vacuum superinsulation and mass flow through the valves and safety disks. Even if the design of the helium relief system does not foresee direct helium venting into the environment, an occasional emergency helium spill may happen. Helium propagation in the atmosphere and the origins of oxygen-deficiency hazards will be discussed.

  18. Fast breeder reactor protection system

    DOEpatents

    van Erp, J.B.

    1973-10-01

    Reactor protection is provided for a liquid-metal-fast breeder reactor core by measuring the coolant outflow temperature from each of the subassemblies of the core. The outputs of the temperature sensors from a subassembly region of the core containing a plurality of subassemblies are combined in a logic circuit which develops a scram alarm if a predetermined number of the sensors indicate an over temperature condition. The coolant outflow from a single subassembly can be mixed with the coolant outflow from adjacent subassemblies prior to the temperature sensing to increase the sensitivity of the protection system to a single subassembly failure. Coherence between the sensors can be required to discriminate against noise signals. (Official Gazette)

  19. POWER GENERATING NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Vernon, H.C.

    1958-03-01

    This patent relates to reactor systems of the type wherein the cooiing medium is a liquid which is converted by the heat of the reaction to steam which is conveyed directly to a pnime mover such as a steam turbine driving a generatore after which it is condensed and returred to the coolant circuit. In this design, the reactor core is disposed within a tank for containing either a slurry type fuel or an aggregation of solid fuel elements such as elongated rods submerged in a liquid moderator such as heavy water. The top of the tank is provided with a nozzle which extends into an expansion chamber connected with the upper end of the tank, the coolant being maintained in the expansion chamber at a level above the nozzle and the steam being formed in the expansion chamber.

  20. Recent developments in Topaz-II reactor safety assessments

    SciTech Connect

    Marshall, A.C. )

    1993-01-01

    In December 1991, the Strategic Defense Initiative Organization (SDIO) decided to investigate the possibility of a US launch of a Russian Topaz-II space nuclear power system. The primary mission goal would be to demonstrate and evaluate nuclear electric propulsion technology to establish a capability for future civilian and military missions. A preliminary nuclear safety analysis was initiated to determine whether or not a space mission could be conducted safely and within budget constraints. This paper describes preliminary safety analysis results and the nuclear safety program now being established for the NEP space test (NEPST).

  1. Simulating experimental investigation on the safety of nuclear heating reactor in loss-of-coolant accidents

    NASA Astrophysics Data System (ADS)

    Xu, Zhanjie

    1996-12-01

    The 5MW low temperature nuclear heating reactor (NHR-5) is a new and advanced type of nuclear reactor developed by Institute of Nuclear Energy Technology (INET) of Tsinghua University of China in 1989. Its main loop is a thermal-hydraulic system with natural circulation. This paper studies the safety of NHR under the condition of loss-of-coolant accidents (LOCAs) by means of simulant experiments. First, the background and necessity of the experiments are presented, then the experimental system, including the thermal-hydraulic system and the data collection system, and similarity criteria are introduced. Up to now, the discharge experiments with the residual heating power (20% rated heating power) have been carried out on the experimental system. The system parameters including circulation flow rate, system pressure, system temperature, void fraction, discharge mass and so on have been recorded and analyzed. Based on the results of the experiments, the conclusions are shown as follos: on the whole, the reactor is safe under the condition of LOCAs, but the thermal vacillations resulting from the vibration of the circulation flow rate are disadvantageous to the internal parts of the reactor core.

  2. Final report-passive safety optimization in liquid sodium-cooled reactors.

    SciTech Connect

    Cahalana, J. E.; Hahn, D.; Nuclear Engineering Division; Korea Atomic Energy Research Inst.

    2007-08-13

    Implications of Advanced Technology Power Conversion and Design Innovations and Simplifications: Investigations of supercritical CO{sub 2} gas turbine Brayton cycles coupled to the sodium-cooled reactors and innovative concepts for sodium-to-CO{sub 2} heat exchangers were performed to discover new designs for high efficiency electricity production. The objective of the analyses was to characterize the design and safety performance of equipment needed to implement the new power cycle. The project included considerations of heat transfer and power conversion systems arrangements and evaluations of systems performance. Task 4--Post Accident Heat Removal and In-Vessel Retention: Test plans were developed to evaluate (1) freezing and plugging of molten metallic fuel in subassembly geometry, (2) retention of metallic fuel core melt debris within reactor vessel structures, and (3) consequences of intermixing of high pressure CO{sub 2} and sodium. The objective of the test plan development was to provide planning for measurements of data needed to characterize the consequences of very low probability accident sequences unique to metallic fuel and CO{sub 2} Brayton power cycles. The project produced three test plans ready for execution.

  3. Liquid metal cooled nuclear reactor plant system

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1993-01-01

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

  4. Safety Design Strategy for the Advanced Test Reactor Diesel Bus (E-3) and Switchgear Replacement Project

    SciTech Connect

    Noel Duckwitz

    2011-06-01

    In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

  5. Safety Design Strategy for the Advanced Test Reactor Primary Coolant Pump and Motor Replacement Project

    SciTech Connect

    Noel Duckwitz

    2011-06-01

    In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

  6. 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.

  7. Safety Characteristics of LBE Cooled Long-Life Small Reactor, 'LSPR'

    SciTech Connect

    Hiroshi Sekimoto; Shinichi Makino

    2002-07-01

    Lead bismuth eutectic (LBE) shows a good performance on neutron economy, and LBE cooled fast reactor can be designed as an excellent long-life small reactor. LBE is good not only for neutron economy but for chemical inertness and high boiling point, which may realize a much safer reactor than conventional sodium-cooled reactor. We have designed such a long-life small reactor and name it LSPR. This paper presents safety characteristics of LSPR. (authors)

  8. Thermal Reactor Code System for Reactor Design and Analysis.

    SciTech Connect

    SUZUKI, TADAKAZU

    2003-04-21

    Version: 00 SRAC95 is a general purpose neutronics code system applicable to core analyses of various types of reactors, including cell calculation with burn up, core calculation for any type of thermal reactor; where core burn up calculation and fuel management were done by an auxiliary code. Since the publication of JAERI-1302 for the revised SRAC in 1986, a number of additions and modifications were made for nuclear data libraries and programs. In this version, many new functions and data are implemented to support nuclear design studies of advanced reactors. SRAC95 can be used for burnup credit analysis within the ORIGEN2 and SWAT (CCC-714) code system.

  9. System safety management lessons learned

    SciTech Connect

    Piatt, J.A.

    1989-05-01

    The Assistant Secretary of the Army for Research, Development and Acquisition directed the Army Safety Center to provide an audit of the causes of accidents and safety of use restrictions on recently fielded systems by tracking residual hazards back through the acquisition process. The objective was to develop ''lessons learned'' that could be applied to the acquisition process to minimize mishaps in fielded systems. System safety management lessons learned are defined as Army practices or policies, derived from past successes and failures, that are expected to be effective in eliminating or reducing specific systemic causes of residual hazards. They are broadly applicable and supportive of the Army structure and acquisition objectives. 29 refs., 7 figs.

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

    SciTech Connect

    Corradini, M. L.

    2015-06-01

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

  11. 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…

  12. Safety of evolutionary and innovative nuclear reactors: IAEA activities and world efforts

    SciTech Connect

    Saito, T.; Gasparini, M.

    2004-07-01

    'Defence in Depth' approach constitutes the basis of the IAEA safety standards for nuclear power plants. Lessons learned from the current generation of reactors suggest that, for the next generation of reactor designs, the Defence in Depth philosophy should be retained, and that its implementation should be guided by the probabilistic insights. Recent developments in the area of general safety requirements based on Defence in Depth approach are examined and summarized. Global efforts to harmonize safety requirements for evolutionary nuclear power plants have involved many countries and organizations such as IAEA, US EPRI and European Utility EUR Organization. In recent years, developments of innovative nuclear power plants are also being discussed. The IAEA is currently developing a safety approach specifically for innovative nuclear reactors. This approach will eventually lead to a proposal of safety requirements for innovative reactors. Such activities related to safety requirements of evolutionary and innovative reactors are introduced. Various evolutionary and innovative reactor designs are reported in the world. The safety design features of evolutionary large LWRs, innovative LWRs, Modular High Temperature Gas Reactors and Small Liquid Metal Cooled LMRs are also introduced. Enhanced safety features proposed in such reactors are discussed and summarized according to the levels of Defence in Depth. For future nuclear plants, international cooperation and harmonization, especially in the area of safety, appear to be inevitable. Based on the past experience with many member states, the IAEA believes itself to be the uniquely positioned international organization to play this key role. (authors)

  13. CONTROL SYSTEM FOR NEUTRONIC REACTORS

    DOEpatents

    Crever, F.E.

    1962-05-01

    BS>A slow-acting shim rod for control of major variations in reactor neutron flux and a fast-acting control rod to correct minor flux variations are employed to provide a sensitive, accurate control system. The fast-acting rod is responsive to an error signal which is produced by changes in the neutron flux from a predetermined optimum level. When the fast rod is thus actuated in a given direction, means is provided to actuate the slow-moving rod in that direction to return the fast rod to a position near the midpoint of its control range. (AEC)

  14. Sodium fast reactor gaps analysis of computer codes and models for accident analysis and reactor safety.

    SciTech Connect

    Carbajo, Juan; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Schmidt, Rodney Cannon; Thomas, Justin; Wei, Tom; Sofu, Tanju; Ludewig, Hans; Tobita, Yoshiharu; Ohshima, Hiroyuki; Serre, Frederic

    2011-06-01

    This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the

  15. NASA aviation safety reporting system

    NASA Technical Reports Server (NTRS)

    Billings, C. E.; Lauber, J. K.; Funkhouser, H.; Lyman, E. G.; Huff, E. M.

    1976-01-01

    The origins and development of the NASA Aviation Safety Reporting System (ASRS) are briefly reviewed. The results of the first quarter's activity are summarized and discussed. Examples are given of bulletins describing potential air safety hazards, and the disposition of these bulletins. During the first quarter of operation, the ASRS received 1464 reports; 1407 provided data relevant to air safety. All reports are being processed for entry into the ASRS data base. During the reporting period, 130 alert bulletins describing possible problems in the aviation system were generated and disseminated. Responses were received from FAA and others regarding 108 of the alert bulletins. Action was being taken with respect to 70 of the 108 responses received. Further studies are planned of a number of areas, including human factors problems related to automation of the ground and airborne portions of the national aviation system.

  16. Incipient Transient Detection in Reactor Systems: Experimental and Theoretical Investigation

    SciTech Connect

    Lefteri H. Tsoukalas; S.T. Revankar; X Wang; R. Sattuluri

    2005-09-27

    The main goal of this research was to develop a method for detecting reactor system transients at the earliest possible time through a comprehensive experimental, testing and benchmarking program. This approach holds strong promise for developing new diagnostic technologies that are non-intrusive, generic and highly portable across different systems. It will help in the design of new generation nuclear power reactors, which utilize passive safety systems with a reliable and non-intrusive multiphase flow diagnostic system to monitor the function of the passive safety systems. The main objective of this research was to develop an improved fuzzy logic based detection method based on a comprehensive experimental testing program to detect reactor transients at the earliest possible time, practically at their birth moment. A fuzzy logic and neural network based transient identification methodology and implemented in a computer code called PROTREN was considered in this research and was compared with SPRT (Sequentially Probability Ratio Testing) decision and Bayesian inference. The project involved experiment, theoretical modeling and a thermal-hydraulic code assessment. It involved graduate and undergraduate students participation providing them with exposure and training in advanced reactor concepts and safety systems. In this final report, main tasks performed during the project period are summarized and the selected results are presented. Detailed descriptions for the tasks and the results are presented in previous yearly reports (Revankar et al 2003 and Revankar et al 2004).

  17. Progress in reliability of fast reactor operation and new trends to increased inherent safety

    SciTech Connect

    Merk, Bruno; Stanculescu, Alexander; Chellapandi, Perumal; Hill, Robert

    2015-06-01

    The reasons for the renewed interest in fast reactors and an overview of the progress in sodium cooled fast reactor operation in the last ten years are given. The excellent operational performance of sodium cooled fast reactors in this period is highlighted as a sound basis for the development of new fast reactors. The operational performance of the BN-600 is compared and evaluated against the performance of German light water reactors to assess the reliability. The relevance of feedback effects for safe reactor design is described, and a new method for the enhancement of feedback effects in fast reactors is proposed. Experimental reactors demonstrating the inherent safety of advanced sodium cooled fast reactor designs are described and the potential safety improvements resulting from the use of fine distributed moderating material are discussed.

  18. Very high temperature measurements: Application to nuclear reactor safety tests

    NASA Astrophysics Data System (ADS)

    Parga, Clemente Jose

    This PhD dissertation focuses on the improvement of very high temperature thermometry (1100ºC to 2480ºC), with special emphasis on the application to the field of nuclear reactor safety and severe accident research. Two main projects were undertaken to achieve this objective: -The development, testing and transposition of high-temperature fixed point (HTFP) metal-carbon eutectic cells, from metrology laboratory precision (+/-0.001ºC) to applied research with a reasonable degradation of uncertainties (+/-3-5ºC). -The corrosion study and metallurgical characterization of Type-C thermocouple (service temp. 2300ºC) prospective sheath material was undertaken to extend the survivability of TCs used for molten metallic/oxide corium thermometry (below 2000ºC).

  19. Nuclear reactor safety research since three mile island.

    PubMed

    Mynatt, F R

    1982-04-01

    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. PMID:17736229

  20. NASA aviation safety reporting system

    NASA Technical Reports Server (NTRS)

    1977-01-01

    During the third quarter of operation of the Aviation Safety Reporting System (ASRS), 1429 reports concerning aviation safety were received from pilots, air traffic controllers, and others in the national aviation system. Details of the administration and results of the program are discussed. The design and construction of the ASRS data base are briefly presented. Altitude deviations and potential aircraft conflicts associated with misunderstood clearances were studied and the results are discussed. Summary data regarding alert bulletins, examples of alert bulletins and responses to them, and a sample of deidentified ASRS reports are provided.

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

    SciTech Connect

    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 reactor concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs.

  2. Water Reactor Safety Research Division quarterly progress report, January 1-March 31, 1980

    SciTech Connect

    Romano, A.J.

    1980-06-01

    The Water Reactor Safety Research Programs Quarterly Report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: LWR Thermal Hydraulic Development, Advanced Code Evaluation, TRAC Code Assessment, and Stress Corrosion Cracking of PWR Steam Generator Tubing.

  3. Water Reactor Safety Research Division. Quarterly progress report, April 1-June 30, 1980

    SciTech Connect

    Abuaf, N.; Levine, M.M.; Saha, P.; van Rooyen, D.

    1980-08-01

    The Water Reactor Safety Research Programs quarterly report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: LWR Thermal Hydraulic Development, Advanced Code Evlauation, TRAC Code Assessment, and Stress Corrosion Cracking of PWR Steam Generator Tubing.

  4. SACRD: a data base for fast reactor safety computer codes, operational procedures

    SciTech Connect

    Forsberg, V.M.; Arwood, J.W.; Greene, N.M.; Raiford, G.B.

    1980-09-01

    SACRD (Safety Analysis Computerized Reactor Data) is a data base of nondesign-related information used in computer codes for fast reactor safety analyses. This document reports the procedures used in SACRD to help assure a reasonable level of integrity of the material contained in the data base. It also serves to document much of the computer software used with the data base.

  5. Shutdown system for a nuclear reactor

    DOEpatents

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

    1984-06-05

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

  6. Shutdown system for a nuclear reactor

    DOEpatents

    Groh, Edward F.; Olson, Arne P.; Wade, David C.; Robinson, Bryan W.

    1984-01-01

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

  7. Experimental and design experience with passive safety features of liquid metal reactors

    SciTech Connect

    Lucoff, D.M.; Waltar, A.E.; Sackett, J.I.; Salvatores, M.; Aizawa, K.

    1992-10-01

    Liquid metal cooled reactors (LMRs) have already been demonstrated to be robust machines. Many reactor designers now believe that it is possible to include in this technology sufficient passive safety that LMRs would be able to survive loss of flow, loss of heat sink, and transient overpower events, even if the plant protective system fails completely and do so without damage to the core. Early whole-core testing in Rapsodie, EBR-II. and FFTF indicate such designs may be possible. The operational safety testing program in EBR-II is demonstrating benign response of the reactor to a full range of controls failures. But additional testing is needed if transient core structural response under major accident conditions is to be properly understood. The proposed international Phase IIB passive safety tests in FFTF, being designed with a particular emphasis on providing, data to understand core bowing extremes, and further tests planned in EBR-II with processed IFR fuel should provide a substantial and unique database for validating the computer codes being used to simulate postulated accident conditions.

  8. Experimental and design experience with passive safety features of liquid metal reactors

    SciTech Connect

    Lucoff, D.M.; Waltar, A.E. ); Sackett, J.I. ); Aizawa, K. )

    1992-07-01

    Liquid metal cooled reactors (LMRs) have already been demonstrated to be robust machines. Many reactor designers now believe that it is possible to include in this technology sufficient passive safety that LMRs would be able to survive loss of flow, loss of heat sink, and transient overpower events, even if the plant protective system fails completely--and do so without damage to the core. Early whole-core testing in Rapsodie, EBR-II, and FFTF indicate such designs may be possible. The operational safety testing program in EBR-II is demonstrating benign response of the reactor to a full range on controls failures. But additional testing is needed if transient core structural response under major accident conditions is to be properly understood. The proposed international Phase IIB passive safety tests in FFTF, being designed with a particular emphasis on providing data to understand core bowing extremes, and further tests planned in EBR-II with processed IFR fuel should provide a substantial and unique database for validating the computer codes being used to simulate postulated accident conditions.

  9. 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.

  10. Physics of reactor safety. Quarterly report, July-September 1980. Volume III

    SciTech Connect

    Not Available

    1980-11-01

    This Quarterly progress report summarizes work done during the months of July-September 1980 in Argonne National Laboratory's Applied Physics and Components Technology Divisions for the Division of Reactor Safety Research of the US Nuclear Regulatory Commission. The work in the Applied Physics Division includes reports on reactor safety modeling and assessment by members of the Reactor Safety Appraisals Section. Work on reactor core thermal-hydraulics is performed in ANL's Components Technology Division, emphasizing 3-dimensional code development for LMFBR accidents under natural convection conditions. An executive summary is provided including a statement of the findings and recommendations of the report.

  11. 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.

  12. NASA aviation safety reporting system

    NASA Technical Reports Server (NTRS)

    1976-01-01

    During the second quarter of the Aviation Safety Reporting System (ASRS) operation, 1,497 reports were received from pilots, controllers, and others in the national aviation system. Details of the administration and results of the program to date are presented. Examples of alert bulletins disseminated to the aviation community are presented together with responses to those bulletins. Several reports received by ASRS are also presented to illustrate the diversity of topics covered by reports to the system.

  13. Reactor safety research programs. Quarterly report, July-September 1983

    SciTech Connect

    Edler, S.K.

    1984-04-01

    Evaluations of nondestructive examination (NDE) techniques and instrumentation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, and examining NDE reliability and probabilistic fracture mechanics. Accelerated pellet-cladding interaction modeling is being conducted to predict the probability of fuel rod failure under normal operating conditions. Experimental data and analytical models are being provided to aid in decision making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Experimental data and validated models are being used to determine a method for evaluating the acceptance of welded or weld-repaired stainless steel piping. Thermal-hydraulic models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. High-temperature materials property tests are being conducted to provide data on severe core damage fuel behavior. Severe fuel damage accident tests are being conducted at the NRU reactor, Chalk River, Canada; and an instrumented fuel assembly irradiation program is being performed at Halden, Norway. Fuel assemblies and analytical support are being provided for experimental programs at other facilities, including the Super Sara Test Program, Ispra, Italy, and experimental programs at the Power Burst Facility.

  14. Reactor safety research programs. Quarterly report, April-June 1982

    SciTech Connect

    Edler, S.K.

    1982-11-01

    This document summarizes work performed by Pacific Northwest Laboratory (PNL) from April 1 through June 30, 1982, for the Division of Accident Evaluation and the Division of Engineering Technology, US Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities.

  15. Reactor safety research programs. Quarterly report, January-March 1982

    SciTech Connect

    Edler, S.K.

    1982-07-01

    This document summarizes work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1982, for the Division of Accident Evaluation and the Division of Engineering Technology, US Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities.

  16. 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...

  17. 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...

  18. NASA aviation safety reporting system

    NASA Technical Reports Server (NTRS)

    1981-01-01

    Aviation safety reports that relate to loss of control in flight, problems that occur as a result of similar sounding alphanumerics, and pilot incapacitation are presented. Problems related to the go around maneuver in air carrier operations, and bulletins (and FAA responses to them) that pertain to air traffic control systems and procedures are included.

  19. Software Safety Risk in Legacy Safety-Critical Computer Systems

    NASA Technical Reports Server (NTRS)

    Hill, Janice; Baggs, Rhoda

    2007-01-01

    Safety-critical computer systems must be engineered to meet system and software safety requirements. For legacy safety-critical computer systems, software safety requirements may not have been formally specified during development. When process-oriented software safety requirements are levied on a legacy system after the fact, where software development artifacts don't exist or are incomplete, the question becomes 'how can this be done?' The risks associated with only meeting certain software safety requirements in a legacy safety-critical computer system must be addressed should such systems be selected as candidates for reuse. This paper proposes a method for ascertaining formally, a software safety risk assessment, that provides measurements for software safety for legacy systems which may or may not have a suite of software engineering documentation that is now normally required. It relies upon the NASA Software Safety Standard, risk assessment methods based upon the Taxonomy-Based Questionnaire, and the application of reverse engineering CASE tools to produce original design documents for legacy systems.

  20. Modification of the Core Cooling System of TRIGA 2000 Reactor

    SciTech Connect

    Umar, Efrizon; Fiantini, Rosalina

    2010-06-22

    To accomplish safety requirements, a set of actions has to be performed following the recommendations of the IAEA safety series 35 applied to research reactor. Such actions are considered in modernization of the old system, improving the core cooling system and safety evaluations. Due to the complexity of the process and the difficulty in putting the apparatus in the reactor core, analytical and experimental study on the determination of flow and temperature distribution in the whole coolant channel are difficult to be done. In the present work, a numerical study of flow and temperature distribution in the coolant channel of TRIGA 2000 has been carried out using CFD package. For this study, simulations were carried out on 3-D tested model. The model consists of the reactor tank, thermal and thermalizing column, reflector, rotary specimen rack, chimney, fuel element, primary pipe, diffuser, beam tube and a part of the core are constructed by 1.50 million unstructured tetrahedral cell elements. The results show that for the initial condition (116 fuel elements in the core) and for the inlet temperature of 24 deg. C and the primary velocity of 5.6 m/s, there no boiling phenomena occur in the coolant channel. Due to this result, it is now possible to improve the core cooling system of TRIGA 2000 reactor. Meanwhile, forced flow from the diffuser system only affected the flow pattern in the outside of chimney and put on a small effect to the fluid flow's velocity in the inside of chimney.

  1. Modification of the Core Cooling System of TRIGA 2000 Reactor

    NASA Astrophysics Data System (ADS)

    Umar, Efrizon; Fiantini, Rosalina

    2010-06-01

    To accomplish safety requirements, a set of actions has to be performed following the recommendations of the IAEA safety series 35 applied to research reactor. Such actions are considered in modernization of the old system, improving the core cooling system and safety evaluations. Due to the complexity of the process and the difficulty in putting the apparatus in the reactor core, analytical and experimental study on the determination of flow and temperature distribution in the whole coolant channel are difficult to be done. In the present work, a numerical study of flow and temperature distribution in the coolant channel of TRIGA 2000 has been carried out using CFD package. For this study, simulations were carried out on 3-D tested model. The model consists of the reactor tank, thermal and thermalizing column, reflector, rotary specimen rack, chimney, fuel element, primary pipe, diffuser, beam tube and a part of the core are constructed by 1.50 million unstructured tetrahedral cell elements. The results show that for the initial condition (116 fuel elements in the core) and for the inlet temperature of 24°C and the primary velocity of 5.6 m/s, there no boiling phenomena occur in the coolant channel. Due to this result, it is now possible to improve the core cooling system of TRIGA 2000 reactor. Meanwhile, forced flow from the diffuser system only affected the flow pattern in the outside of chimney and put on a small effect to the fluid flow's velocity in the inside of chimney.

  2. System safety education focused on industrial engineering

    NASA Technical Reports Server (NTRS)

    Johnston, W. L.; Morris, R. S.

    1971-01-01

    An educational program, designed to train students with the specific skills needed to become safety specialists, is described. The discussion concentrates on application, selection, and utilization of various system safety analytical approaches. Emphasis is also placed on the management of a system safety program, its relationship with other disciplines, and new developments and applications of system safety techniques.

  3. Small reactor power systems for manned planetary surface bases

    NASA Astrophysics Data System (ADS)

    Bloomfield, Harvey S.

    1987-12-01

    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.

  4. 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.

  5. Tandem Mirror Reactor Systems Code (Version I)

    SciTech Connect

    Reid, R.L.; Finn, P.A.; Gohar, M.Y.; Barrett, R.J.; Gorker, G.E.; Spampinaton, P.T.; Bulmer, R.H.; Dorn, D.W.; Perkins, L.J.; Ghose, S.

    1985-09-01

    A computer code was developed to model a Tandem Mirror Reactor. Ths is the first Tandem Mirror Reactor model to couple, in detail, the highly linked physics, magnetics, and neutronic analysis into a single code. This report describes the code architecture, provides a summary description of the modules comprising the code, and includes an example execution of the Tandem Mirror Reactor Systems Code. Results from this code for two sensitivity studies are also included. These studies are: (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power; and (2) to determine the impact of reactor power level on cost.

  6. Comparison of the N Reactor and Ignalina Unit No. 2 Level 1 Probabilistic Safety Assessments

    SciTech Connect

    Coles, G.A.; McKay, S.L.

    1995-06-01

    A multilateral team recently completed a full-scope Level 1 Probabilistic Safety Assessment (PSA) on the Ignalina Unit No. 2 reactor plant in Lithuania. This allows comparison of results to those of the PSA for the U.S. Department of Energy`s (DOE) N Reactor. The N Reactor, although unique as a Western design, has similarities to Eastern European and Soviet graphite block reactors.

  7. An Autonomous Flight Safety System

    NASA Technical Reports Server (NTRS)

    Bull, James B.; Lanzi, Raymond J.

    2007-01-01

    The Autonomous Flight Safety System (AFSS) being developed by NASA s Goddard Space Flight Center s Wallops Flight Facility and Kennedy Space Center has completed two successful developmental flights and is preparing for a third. AFSS has been demonstrated to be a viable architecture for implementation of a completely vehicle based system capable of protecting life and property in event of an errant vehicle by terminating the flight or initiating other actions. It is capable of replacing current human-in-the-loop systems or acting in parallel with them. AFSS is configured prior to flight in accordance with a specific rule set agreed upon by the range safety authority and the user to protect the public and assure mission success. This paper discusses the motivation for the project, describes the method of development, and presents an overview of the evolving architecture and the current status.

  8. Reference Reactor Module for the Affordable Fission Surface Power System

    SciTech Connect

    Poston, David I.; Kapernick, Richard J.; Dixon, David D.; Amiri, Benjamin W.; Marcille, Thomas F.

    2008-01-21

    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The requirements of many surface power applications allow the consideration of systems with much less development risk than most other space reactor applications, because of modest power (10s of kWe) and no driving need for minimal mass (allowing temperatures <1000 K). The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. This paper describes the reference AFSPS reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based, UO{sub 2}-fueled, liquid metal-cooled fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. One of the important 'affordability' attributes is that the concept has been designed to minimize both the technical and programmatic safety risk.

  9. Nuclear electric propulsion reactor control systems status

    NASA Technical Reports Server (NTRS)

    Ferg, D. A.

    1973-01-01

    The thermionic reactor control system design studies conducted over the past several years for a nuclear electric propulsion system are described and summarized. The relevant reactor control system studies are discussed in qualitative terms, pointing out the significant advantages and disadvantages including the impact that the various control systems would have on the nuclear electric propulsion system design. A recommendation for the reference control system is made, and a program for future work leading to an engineering model is described.

  10. The LOCA performance of the AP600 passive safety systems

    SciTech Connect

    Kemper, R.M.; Hochreiter, L.E.; Takeuchi, K.; Garner, D.C.; Nguyen, S.B.; Cunningham, J.P. ); Lee, S.N.K.; Tehrani, A.A.K.; Yang, H.; Bratby, P.A.W. )

    1992-01-01

    The AP600 is an advanced passive safeguards pressurized water reactor (PWR) that is being developed jointly by Westinghouse Electric Corporation, the U.S. Department of Energy, and the Electrical Power Research Institute. The plant has a thermal rating of 1940 MW (thermal) [600 MW(electric)] and has been designed with passive safeguard systems that utilize gravity feed injection rather than safety-grade active pumps and equipment. Calculations performed for a range of break sizes used locations to find the worst set of conditions for depressurizing the reactor coolant system. The main criterion was system inventory such that the core remained covered. The resulting break spectrum study indicates only that the double-ended guillotine shear of the direct vessel injection line (a .68-in. line that feeds the emergency core coolant flow into the vessel) resulted in a momentary core uncover. For all other small-break cases, the core remained covered as the reactor coolant system depressurized. The passive safety systems provided sufficient mass flow to the reactor vessel such that even under the more conservative Appendix K assumptions, the core remained covered and in a coolable state. The LOCA analysis performed for the AP600 confirms that passive safety systems can provide the core cooling necessary to meet the requirements of 10CFR50.46 with ample margin.

  11. Autonomous Control of Space Reactor Systems

    SciTech Connect

    Belle R. Upadhyaya; K. Zhao; S.R.P. Perillo; Xiaojia Xu; M.G. Na

    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.

  12. Guidelines for preparing criticality safety evaluations at Department of Energy non-reactor nuclear facilities

    SciTech Connect

    Not Available

    1993-11-01

    This document contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE non-reactor nuclear facilities. Adherence to these guidelines will provide consistency and uniformity in criticality safety evaluations (CSEs) across the complex and will document compliance with the requirements of DOE Order 5480.24.

  13. CONCEPTUAL DESIGN OF A LUNAR REGOLITH CLUSTERED-REACTOR SYSTEM

    SciTech Connect

    John Darrell Bess

    2009-06-01

    It is proposed that a fast-fission, heatpipe-cooled, lunar-surface power reactor system be divided into subcritical units that could be launched safely without the incorporation of additional spectral shift absorbers or other complex means of control. The reactor subunits are to be emplaced directly into the lunar regolith utilizing the regolith not just for shielding but as the reflector material to increase the neutron economy of the system. While a single subunit cannot achieve criticality by itself, coordinated placement of additional subunits will provide a critical reactor system for lunar surface power generation. A lunar regolith clustered-reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of a slight increase in launch mass per rated power level and an overall reduction in neutron economy when compared to a single-reactor system. Additional subunits may be launched with future missions to increase the cluster size and power according to desired lunar base power demand and lifetime. The results address the potential uncertainties associated with the lunar regolith material and emplacement of the subunit systems. Physical distance between subunits within the clustered emplacement exhibits the most significant feedback regarding changes in overall system reactivity. Narrow, deep holes will be the most effective in reducing axial neutron leakage from the core. The variation in iron concentration in the lunar regolith can directly influence the overall system reactivity although its effects are less than the more dominant factors of subunit emplacement.

  14. Thermal Reactor Code System for Reactor Design and Analysis.

    Energy Science and Technology Software Center (ESTSC)

    2003-04-21

    Version: 00 SRAC95 is a general purpose neutronics code system applicable to core analyses of various types of reactors, including cell calculation with burn up, core calculation for any type of thermal reactor; where core burn up calculation and fuel management were done by an auxiliary code. Since the publication of JAERI-1302 for the revised SRAC in 1986, a number of additions and modifications were made for nuclear data libraries and programs. In this version,more » many new functions and data are implemented to support nuclear design studies of advanced reactors. SRAC95 can be used for burnup credit analysis within the ORIGEN2 and SWAT (CCC-714) code system.« less

  15. The 5-kwe reactor thermoelectric system summary

    NASA Technical Reports Server (NTRS)

    Vanosdol, J. H. (Editor)

    1973-01-01

    Design of the 5-kwe reactor thermoelectric system was initiated in February 1972 and extended through the conceptual design phase into the preliminary design phase. Design effort was terminated in January, 1973. This report documents the system and component requirements, design approaches, and performance and design characteristics for the 5-kwe system. Included is summary information on the reactor, radiation shields, power conversion systems, thermoelectric pump, radiator/structure, liquid metal components, and the control system.

  16. SP-100 Space Reactor Power System readiness

    NASA Astrophysics Data System (ADS)

    Josloff, A. T.; Matteo, D. N.; Bailey, H. S.

    The SP-100 Space Reactor Power System is being developed by GE, under contract to the U.S. Department of Energy, to provide electrical power in the range of 10's to 100's of kW. The system represents an enabling technology for a wide variety of earth orbital and interplanetary science missions, nuclear electric propulsion (NEP) stages, and lunar/Mars surface power for the Space Exploration Initiative (SEI). An effective infrastructure of Industry, National Laboratories and Government agencies has made substantial progress since the 1988 System Design Review. Hardware development and testing has progressed to the point of resolving all key technical feasibility issues. The technology and design is now at a state of readiness to support the definition of early flight demonstration missions. Of particular importance is that SP-100 meets the demanding U.S. safety, performance, reliability and life requirements. The system is scalable and flexible and can be configured to provide 10's to 100's of kWe without repeating development work and can meet DoD goals for an early, low-power demonstration flight in the 1996 - 1997 time frame.

  17. Fission control system for nuclear reactor

    DOEpatents

    Conley, G.H.; Estes, G.P.

    Control system for nuclear reactor comprises a first set of reactivity modifying rods fixed in a reactor core with their upper ends stepped in height across the core, and a second set of reactivity modifying rods movable vertically within the reactor core and having their lower ends stepped to correspond with the stepped arrangement of the first set of rods, pairs of the rods of the first and second sets being in coaxial alignment.

  18. Small reactor power system for space application

    NASA Technical Reports Server (NTRS)

    Shirbacheh, M.

    1987-01-01

    A development history and comparative performance capability evaluation is presented for spacecraft nuclear powerplant Small Reactor Power System alternatives. The choice of power conversion technology depends on the reactor's operating temperature; thermionic, thermoelectric, organic Rankine, and Alkali metal thermoelectric conversion are the primary power conversion subsystem technology alternatives. A tabulation is presented for such spacecraft nuclear reactor test histories as those of SNAP-10A, SP-100, and NERVA.

  19. Inherent safety of minimum-burnup breed and burn reactors

    SciTech Connect

    Qvist, S.; Reenspan, E.

    2012-07-01

    Reactors that aim to sustain the breed and burn (B and B) mode of operation at minimum discharge burnup require excellent neutron economy, Minimum-burnup B and B cores are generally large and feature low neutron leakage probability and a hard neutron spectrum. While highly promising fuel cycles can be achieved with such designs, the very same features are pushing the limits of the core's ability to passively respond safely to unprotected accidents. Low leakage minimum-burnup sodium-cooled B and B cores have a large positive coolant void-worth and coolant temperature reactivity coefficient. In this study, the applicability of major approaches for fast reactor void-worth reduction is evaluated specifically for B and B cores. The design, shuffling scheme and performance of a new metallic-fueled, sodium-cooled minimum burnup B and B core, used as basis for the void-worth reduction analysis, is presented. The analysis shows that reactivity control systems based on passive {sup 6}Li injection during temperature excursions are the only option able to provide negative void-worth without significantly increasing the minimum burnup required for sustaining the B and B mode of operation. A new type of lithium expansion module (LEM) system was developed specifically for B and B cores and its effect on core performance is presented. (authors)

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

    SciTech Connect

    Curtis Smith; Cristian Rabiti; Richard Martineau

    2012-11-01

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

  1. Development and methodology of level 1 probability safety assessment at PUSPATI TRIGA Reactor

    SciTech Connect

    Maskin, Mazleha; Tom, Phongsakorn Prak; Lanyau, Tonny Anak; Saad, Mohamad Fauzi; Ismail, Ahmad Razali; Abu, Mohamad Puad Haji; Brayon, Fedrick Charlie Matthew; Mohamed, Faizal

    2014-02-12

    As a consequence of the accident at the Fukushima Dai-ichi Nuclear Power Plant in Japan, the safety aspects of the one and only research reactor (31 years old) in Malaysia need be reviewed. Based on this decision, Malaysian Nuclear Agency in collaboration with Atomic Energy Licensing Board and Universiti Kebangsaan Malaysia develop a Level-1 Probability Safety Assessment on this research reactor. This work is aimed to evaluate the potential risks of incidents in RTP and at the same time to identify internal and external hazard that may cause any extreme initiating events. This report documents the methodology in developing a Level 1 PSA performed for the RTP as a complementary approach to deterministic safety analysis both in neutronics and thermal hydraulics. This Level-1 PSA work has been performed according to the procedures suggested in relevant IAEA publications and at the same time numbers of procedures has been developed as part of an Integrated Management System programme implemented in Nuclear Malaysia.

  2. Development and methodology of level 1 probability safety assessment at PUSPATI TRIGA Reactor

    NASA Astrophysics Data System (ADS)

    Maskin, Mazleha; Tom, Phongsakorn Prak; Lanyau, Tonny Anak; Brayon, Fedrick Charlie Matthew; Mohamed, Faizal; Saad, Mohamad Fauzi; Ismail, Ahmad Razali; Abu, Mohamad Puad Haji

    2014-02-01

    As a consequence of the accident at the Fukushima Dai-ichi Nuclear Power Plant in Japan, the safety aspects of the one and only research reactor (31 years old) in Malaysia need be reviewed. Based on this decision, Malaysian Nuclear Agency in collaboration with Atomic Energy Licensing Board and Universiti Kebangsaan Malaysia develop a Level-1 Probability Safety Assessment on this research reactor. This work is aimed to evaluate the potential risks of incidents in RTP and at the same time to identify internal and external hazard that may cause any extreme initiating events. This report documents the methodology in developing a Level 1 PSA performed for the RTP as a complementary approach to deterministic safety analysis both in neutronics and thermal hydraulics. This Level-1 PSA work has been performed according to the procedures suggested in relevant IAEA publications and at the same time numbers of procedures has been developed as part of an Integrated Management System programme implemented in Nuclear Malaysia.

  3. Fast reactor safety: proceedings of the international topical meeting. Volume 1

    SciTech Connect

    Not Available

    1985-07-01

    The emphasis of this meeting was on the safety-related aspects of fast reactor design, analysis, licensing, construction, and operation. Relative to past meetings, there was less emphasis on the scientific and technological basis for accident assessment. Because of its broad scope, the meeting attracted 217 attendees from a wide cross section of the design, safety analysis, and safety technology communities. Eight countries and two international organizations were represented. A total of 126 papers were presented, with contributions from the United States, France, Japan, the United Kingdom, Germany, and Italy. Sessions covered in Volume 1 include: impact of safety and licensing considerations on fast reactor design; safety aspects of innovative designs; intra-subassembly behavior; operational safety; design accommodation of seismic and other external events; natural circulation; safety design concepts; safety implications derived from operational plant data; decay heat removal; and assessment of HCDA consequences.

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

    SciTech Connect

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

    1985-09-01

    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 design 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.

  5. A study on reactor core failure thresholds to safety operation of LMFBR

    SciTech Connect

    Kazuo, Haga; Hiroshi, Endo; Tomoko, Ishizu; Yoshihisa, Shindo

    2006-07-01

    Japan Nuclear Safety Organization (JNES) has been developing the methodology and computer codes for applying level-1 PSA to LMFBR. Many of our efforts have been directed to the judging conditions of reactor core damage and the time allowed to initiate the accident management. Several candidates of the reactor core failure threshold were examined to a typical proto-type LMFBR with MOX fuel based on the plant thermal-hydraulic analyses to the actual progressions leading to the core damage. The results of the present study showed that the judging condition of coolant-boundary integrity failure, 750 degree-C of the boundary temperature, is enough as the threshold of core damage to PLOHS (protected loss-of-heat sink). High-temperature fuel cladding creep failure will not take place before the coolant-boundary reaches the judging temperature and sodium boiling will not occur due to the system pressure rise. In cases of ATWS (anticipated transient without scrum) the accident progression is so fast and the reactor core damage will be inevitable even a realistic negative reactivity insertion due to the temperature rise is considered. Only in the case of ULOHS (unprotected loss-of-heat sink) a relatively long time of 11 min will be allowed till the shut-down of the reactor before the core damage. (authors)

  6. 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.

  7. Self-actuating reactor shutdown system

    DOEpatents

    Barrus, Donald M.; Brummond, Willian A; Peterson, Leslie F.

    1988-01-01

    A control system for the automatic or self-actuated shutdown or "scram" of a nuclear reactor. The system is capable of initiating scram insertion by a signal from the plant protection system or by independent action directly sensing reactor conditions of low-flow or over-power. Self-actuation due to a loss of reactor coolant flow results from a decrease of pressure differential between the upper and lower ends of an absorber element. When the force due to this differential falls below the weight of the element, the element will fall by gravitational force to scram the reactor. Self-actuation due to high neutron flux is accomplished via a valve controlled by an electromagnet and a thermionic diode. In a reactor over-power, the diode will be heated to a change of state causing the electromagnet to be shorted thereby actuating the valve which provides the changed flow and pressure conditions required for scramming the absorber element.

  8. TREAT Reactor Control and Protection System

    SciTech Connect

    Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.; Lenkszus, F.R.; McDowell, W.P.

    1985-01-01

    The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS). The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab.

  9. Safety Analyses at the Idaho National Engineering and Environmental Laboratory Test Reactor Area - Past to Present

    SciTech Connect

    Ambrosek, Richard Garry; Ingram, Frederick William

    1999-11-01

    Test reactors are unique in that the core configuration may change with each operating interval. The process of safety analyses for test reactors at the Idaho National Engineering and Environmental Test Reactor Area has evolved as the computing capabilities, software, and regulatory requirements have changed. The evaluations for experiments and the reactor have moved from measurements in a set configuration and then application to other configurations with a relatively large error to modeling in three-dimensions and explicit analyses for each experiment and operating interval. This evolution is briefly discussed for the Test Reactor Area.

  10. 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. PMID:26123105

  11. Catalytic reactor system for the tritium emissions reduction facility

    SciTech Connect

    Wieneke, R.E.

    1991-12-31

    Two platinum catalyst reactor subsystems have been built for the new Tritium Emissions Reduction Facility (TERF) at Mound. The two parallel subsystems each consist of three major components: a passive conservation heat exchanger, an electric preheater, and a catalytic reactor. All subsystem components and interconnecting piping are fabricated from Inconel 625 for high temperature strength and corrosion resistance. System connections are welded for longevity and reliability. Active elements are backed up by installed spares, and the reactor catalyst is replaceable. Since double containment of tritium processing systems is an important safety concept, the entire subsystem is enclosed in a stainless steel glovebox. Careful planning during the design phase created thermal isolation from the glovebox, and the ability to translate the entire subsystem from the glovebox for major maintenance. 4 refs.

  12. Catalytic reactor system for the tritium emissions reduction facility

    SciTech Connect

    Wieneke, R.E.

    1991-01-01

    Two platinum catalyst reactor subsystems have been built for the new Tritium Emissions Reduction Facility (TERF) at Mound. The two parallel subsystems each consist of three major components: a passive conservation heat exchanger, an electric preheater, and a catalytic reactor. All subsystem components and interconnecting piping are fabricated from Inconel 625 for high temperature strength and corrosion resistance. System connections are welded for longevity and reliability. Active elements are backed up by installed spares, and the reactor catalyst is replaceable. Since double containment of tritium processing systems is an important safety concept, the entire subsystem is enclosed in a stainless steel glovebox. Careful planning during the design phase created thermal isolation from the glovebox, and the ability to translate the entire subsystem from the glovebox for major maintenance. 4 refs.

  13. 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.

  14. 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.

  15. 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.

  16. Slow control systems of the Reactor Experiment for Neutrino Oscillation

    NASA Astrophysics Data System (ADS)

    Choi, J. H.; Jang, H. I.; Choi, W. Q.; Choi, Y.; Jang, J. S.; Jeon, E. J.; Joo, K. K.; Kim, B. R.; Kim, H. S.; Kim, J. Y.; Kim, S. B.; Kim, S. Y.; Kim, W.; Kim, Y. D.; Ko, Y. J.; Lee, J. K.; Lim, I. T.; Pac, M. Y.; Park, I. G.; Park, J. S.; Park, R. G.; Seo, H. K.; Seo, S. H.; Shin, C. D.; Siyeon, K.; Yeo, I. S.; Yu, I.

    2016-02-01

    The RENO experiment has been in operation since August 2011 to measure reactor antineutrino disappearance using identical near and far detectors. For accurate measurements of neutrino mixing parameters and efficient data taking, it is crucial to monitor and control the detector in real time. Environmental conditions also need to be monitored for stable operation of detectors as well as for safety reasons. In this paper, we report the design, hardware, operation, and performance of the slow control system.

  17. Designing visual displays and system models for safe reactor operations

    SciTech Connect

    Brown-VanHoozer, S.A.

    1995-12-31

    The material presented in this paper is based on two studies involving the design of visual displays and the user`s prospective model of a system. The studies involve a methodology known as Neuro-Linguistic Programming and its use in expanding design choices from the operator`s perspective image. The contents of this paper focuses on the studies and how they are applicable to the safety of operating reactors.

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

    SciTech Connect

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

    2002-07-15

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

  19. Code development incorporating environmental, safety, and economic aspects of fusion reactors (FY 89--91)

    SciTech Connect

    Ho, S.K.; Fowler, T.K.; Holdren, J.P.

    1991-11-01

    This report discusses the following aspects of Fusion reactors.: Activation Analysis; Tritium Inventory; Environmental and Safety Indices and Their Graphical Representation; Probabilistic Risk Assessment (PRA) and Decision Analysis; Plasma Burn Control -- Application to ITER; and Other Applications.

  20. Human factors and safety issues associated with actinide retrieval from spent light water reactor fuel assemblies

    SciTech Connect

    Spelt, P.F.

    1992-01-01

    A major problem in environmental restoration and waste management is the disposition of used fuel assemblies from the many light water reactors in the United States, which present a radiation hazard to those whose job is to dispose of them, with a similar threat to the general environment associated with long-term storage in fuel repositories around the country. Actinides resident in the fuel pins as a result of their use in reactor cores constitute a significant component of this hazard. Recently, the Department of Energy has initiated an Actinide Recycle Program to study the feasibility of using pyrochemical (molten salt) processes to recover actinides from the spent fuel assemblies of commercial reactors. This project concerns the application of robotics technology to the operation and maintenance functions of a plant whose objective is to recover actinides from spent fuel assemblies, and to dispose of the resulting hardware and chemical components from this process. Such a procedure involves a number of safety and human factors issues. The purpose of the project is to explore the use of robotics and artificial intelligence to facilitate accomplishment of the program goals while maintaining the safety of the humans doing the work and the integrity of the environment. This project will result in a graphic simulation on a Silicon Graphics workstation as a proof of principle demonstration of the feasibility of using robotics along with an intelligent operator interface. A major component of the operator-system interface is a hybrid artificial intelligence system developed at Oak Ridge National Laboratory, which combines artificial neural networks and an expert system into a hybrid, self-improving computer-based system interface. 10 refs.

  1. Human factors and safety issues associated with actinide retrieval from spent light water reactor fuel assemblies

    SciTech Connect

    Spelt, P.F.

    1992-08-01

    A major problem in environmental restoration and waste management is the disposition of used fuel assemblies from the many light water reactors in the United States, which present a radiation hazard to those whose job is to dispose of them, with a similar threat to the general environment associated with long-term storage in fuel repositories around the country. Actinides resident in the fuel pins as a result of their use in reactor cores constitute a significant component of this hazard. Recently, the Department of Energy has initiated an Actinide Recycle Program to study the feasibility of using pyrochemical (molten salt) processes to recover actinides from the spent fuel assemblies of commercial reactors. This project concerns the application of robotics technology to the operation and maintenance functions of a plant whose objective is to recover actinides from spent fuel assemblies, and to dispose of the resulting hardware and chemical components from this process. Such a procedure involves a number of safety and human factors issues. The purpose of the project is to explore the use of robotics and artificial intelligence to facilitate accomplishment of the program goals while maintaining the safety of the humans doing the work and the integrity of the environment. This project will result in a graphic simulation on a Silicon Graphics workstation as a proof of principle demonstration of the feasibility of using robotics along with an intelligent operator interface. A major component of the operator-system interface is a hybrid artificial intelligence system developed at Oak Ridge National Laboratory, which combines artificial neural networks and an expert system into a hybrid, self-improving computer-based system interface. 10 refs.

  2. Reactor power system/spacecraft integration

    NASA Technical Reports Server (NTRS)

    Elms, R. V.

    1985-01-01

    The new national initiative in space reactor technology evaluation and development is strongly tied to mission applications and to spacecraft and space transportation system (STS) compatibility. This paper discusses the power system integration interfaces with potential using spacecraft and the STS, and the impact of these requirements on the design. The integration areas of interest are mechanical, thermal, electrical, attitude control, and mission environments. The mission environments include space vacuum, solar input, heat sink, space radiation, weapons effects, and reactor power system radiation environments. The natural, reactor, and weapons effects radiation must be evaluated and combined to define the design requirements for spacecraft electronic equipment.

  3. Cooling system for a nuclear reactor

    DOEpatents

    Amtmann, Hans H.

    1982-01-01

    A cooling system for a gas-cooled nuclear reactor is disclosed which includes at least one primary cooling loop adapted to pass coolant gas from the reactor core and an associated steam generator through a duct system having a main circulator therein, and at least one auxiliary cooling loop having communication with the reactor core and adapted to selectively pass coolant gas through an auxiliary heat exchanger and circulator. The main and auxiliary circulators are installed in a common vertical cavity in the reactor vessel, and a common return duct communicates with the reactor core and intersects the common cavity at a junction at which is located a flow diverter valve operative to effect coolant flow through either the primary or auxiliary cooling loops.

  4. SIMPLIFIED SODIUM GRAPHITE REACTOR SYSTEM

    DOEpatents

    Dickinson, R.W.

    1963-03-01

    This patent relates to a nuclear power reactor comprising a reactor vessel, shielding means positioned at the top of said vessel, means sealing said reactor vessel to said shielding means, said vessel containing a quantity of sodium, a core tank, unclad graphite moderator disposed in said tank, means including a plurality of process tubes traversing said tank for isolating said graphite from said sodium, fuel elements positioned in said process tubes, said core tank being supported in spaced relation to the walls and bottom of said reactor vessel and below the level of said sodium, neutron shielding means positioned adjacent said core tank between said core tank and the walls of said vessel, said neutron shielding means defining an annuiar volume adjacent the inside wall of said reactor vessel, inlet plenum means below said core tank for providing a passage between said annular volume and said process tubes, heat exchanger means removably supported from the first-named shielding means and positioned in said annular volume, and means for circulating said sodium over said neutron shielding means down through said heat exchanger, across said inlet plenum and upward through said process tubes, said last-named means including electromagnetic pumps located outside said vessel and supported on said vessel wall between said heat exchanger means and said inlet plenum means. (AEC)

  5. CRYOGENIC UPPER STAGE SYSTEM SAFETY

    NASA Technical Reports Server (NTRS)

    Smith, R. Kenneth; French, James V.; LaRue, Peter F.; Taylor, James L.; Pollard, Kathy (Technical Monitor)

    2005-01-01

    NASA s Exploration Initiative will require development of many new systems or systems of systems. One specific example is that safe, affordable, and reliable upper stage systems to place cargo and crew in stable low earth orbit are urgently required. In this paper, we examine the failure history of previous upper stages with liquid oxygen (LOX)/liquid hydrogen (LH2) propulsion systems. Launch data from 1964 until midyear 2005 are analyzed and presented. This data analysis covers upper stage systems from the Ariane, Centaur, H-IIA, Saturn, and Atlas in addition to other vehicles. Upper stage propulsion system elements have the highest impact on reliability. This paper discusses failure occurrence in all aspects of the operational phases (Le., initial burn, coast, restarts, and trends in failure rates over time). In an effort to understand the likelihood of future failures in flight, we present timelines of engine system failures relevant to initial flight histories. Some evidence suggests that propulsion system failures as a result of design problems occur shortly after initial development of the propulsion system; whereas failures because of manufacturing or assembly processing errors may occur during any phase of the system builds process, This paper also explores the detectability of historical failures. Observations from this review are used to ascertain the potential for increased upper stage reliability given investments in integrated system health management. Based on a clear understanding of the failure and success history of previous efforts by multiple space hardware development groups, the paper will investigate potential improvements that can be realized through application of system safety principles.

  6. Optimized Battery-Type Reactor Primary System Design Utilizing Lead

    SciTech Connect

    Yu, Yong H.; Son, Hyoung M.; Lee, Il S.; Suh, Kune Y.

    2006-07-01

    A number of small and medium size reactors are being developed worldwide as well as large electricity generation reactors for co-generation, district heating or desalination. The Seoul National University has started to develop 23 MWth BORIS (Battery Optimized Reactor Integral System) as a multi-purpose reactor. BORIS is an integral-type optimized fast reactor with an ultra long life core. BORIS is being designed to meet the Generation IV nuclear energy system goals of sustainability, safety, reliability and economics. Major features of BORIS include 20 consecutive years of operation without refueling; elimination of an intermediate heat transport loop and main coolant pump; open core without individual subassemblies; inherent negative reactivity feedback; and inherent load following capability. Its one mission is to provide incremental electricity generation to match the needs of developing nations and especially remote communities without major electrical grid connections. BORIS consists of a reactor module, heat exchanger, coolant module, guard vessel, reactor vessel auxiliary cooling system (RVACS), secondary system, containment and the seismic isolation. BORIS is designed to generate 10 MWe with the resulting thermal efficiency of 45 %. BORIS uses lead as the primary system coolant because of the inherent safety of the material. BORIS is coupled with a supercritical carbon dioxide Brayton cycle as the secondary system to gain a high cycle efficiency in the range of 45 %. The reference core consists of 757 fuel rods without assembly with an active core height of 0.8 m. The BORIS core consists of single enrichment zone composed of a Pu-MA (minor actinides)-U-N fuel and a ferritic-martensitic stainless steel clad. This study is intended to set up appropriate reactor vessel geometry by performing thermal hydraulic analysis on RVACS using computational fluid dynamics codes; to examine the liquid metal coolant behavior along the subchannels; to find out whether the

  7. System Safety Common Cause Analysis

    Energy Science and Technology Software Center (ESTSC)

    1992-03-10

    The COMCAN fault tree analysis codes are designed to analyze complex systems such as nuclear plants for common causes of failure. A common cause event, or common mode failure, is a secondary cause that could contribute to the failure of more than one component and violates the assumption of independence. Analysis of such events is an integral part of system reliability and safety analysis. A significant common cause event is a secondary cause common tomore » all basic events in one or more minimal cut sets. Minimal cut sets containing events from components sharing a common location or a common link are called common cause candidates. Components share a common location if no barrier insulates any one of them from the secondary cause. A common link is a dependency among components which cannot be removed by a physical barrier (e.g.,a common energy source or common maintenance instructions).« less

  8. Westinghouse Small Modular Reactor nuclear steam supply system design

    SciTech Connect

    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 and integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam

  9. Thermionic reactor electric propulsion system requirements.

    NASA Technical Reports Server (NTRS)

    Mondt, J. F.; Sawyer, C. D.; Schaupp, R. W.

    1972-01-01

    Results of mission analysis, system analysis and mission engineering studies to find a single nuclear electric propulsion (NEP) system which would be applicable for a broad range of unmanned outer planet missions. The NEP system studied uses an in-core nuclear thermionic reactor as the electric power source and mercury bombardment ion engines for propulsion. Many requirements, which are imposed on the NEP system by the mission, were determined from the studies in the process of trying to find a single NEP system for many missions. It is concluded that a single thermionic reactor NEP system could be useful for a broad range of unmanned outer planet missions. The thermionic reactor NEP system should have a power level in the range from 70 to 120 kWe, a system specific weight of approximately 30 kg/kWe, and a full power output capability of 20,000 hr.

  10. Scanning tunneling microscope assembly, reactor, and system

    DOEpatents

    Tao, Feng; Salmeron, Miquel; Somorjai, Gabor A

    2014-11-18

    An embodiment of a scanning tunneling microscope (STM) reactor includes a pressure vessel, an STM assembly, and three spring coupling objects. The pressure vessel includes a sealable port, an interior, and an exterior. An embodiment of an STM system includes a vacuum chamber, an STM reactor, and three springs. The three springs couple the STM reactor to the vacuum chamber and are operable to suspend the scanning tunneling microscope reactor within the interior of the vacuum chamber during operation of the STM reactor. An embodiment of an STM assembly includes a coarse displacement arrangement, a piezoelectric fine displacement scanning tube coupled to the coarse displacement arrangement, and a receiver. The piezoelectric fine displacement scanning tube is coupled to the coarse displacement arrangement. The receiver is coupled to the piezoelectric scanning tube and is operable to receive a tip holder, and the tip holder is operable to receive a tip.

  11. Fluidized-Bed Reactor System

    NASA Technical Reports Server (NTRS)

    Morrison, A. D.

    1985-01-01

    Gas pyrolysis in hot fluidized beds minimized by use of selectively filtered radiation and parabolic cavity. Reactor is parabolic cavity of two or more axes in which light emanating from one axis bounces off walls of cavity and passes through object axis to heat sample.

  12. REACTOR CONTROL ROD OPERATING SYSTEM

    DOEpatents

    Miller, G.

    1961-12-12

    A nuclear reactor control rod mechanism is designed which mechanically moves the control rods into and out of the core under normal conditions but rapidly forces the control rods into the core by catapultic action in the event of an emergency. (AEC)

  13. Modeling and performance of the MHTGR (Modular High-Temperature Gas-Cooled Reactor) reactor cavity cooling system

    SciTech Connect

    Conklin, J.C. )

    1990-04-01

    The Reactor Cavity Cooling System (RCCS) of the Modular High- Temperature Gas-Cooled Reactor (MHTGR) proposed by the U.S. Department of Energy is designed to remove the nuclear afterheat passively in the event that neither the heat transport system nor the shutdown cooling circulator subsystem is available. A computer dynamic simulation for the physical and mathematical modeling of and RCCS is described here. Two conclusions can be made form computations performed under the assumption of a uniform reactor vessel temperature. First, the heat transferred across the annulus from the reactor vessel and then to ambient conditions is very dependent on the surface emissivities of the reactor vessel and RCCS panels. These emissivities should be periodically checked to ensure the safety function of the RCCS. Second, the heat transfer from the reactor vessel is reduced by a maximum of 10% by the presence of steam at 1 atm in the reactor cavity annulus for an assumed constant in the transmission of radiant energy across the annulus can be expected to result in an increase in the reactor vessel temperature for the MHTGR. Further investigation of participating radiation media, including small particles, in the reactor cavity annulus is warranted. 26 refs., 7 figs., 1 tab.

  14. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    SciTech Connect

    Neil Todreas; Pavel Hejzlar

    2008-06-30

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

  15. THE RADIATION SAFETY INFORMATION COMPUTATIONAL CENTER: A RESOURCE FOR REACTOR DOSIMETRY SOFTWARE AND NUCLEAR DATA

    SciTech Connect

    Kirk, Bernadette Lugue

    2009-01-01

    The Radiation Safety Information Computational Center (RSICC) was established in 1963 to collect and disseminate computational nuclear technology in the form of radiation transport, shielding and safety software and corresponding nuclear cross sections. Approximately 1700 nuclear software and data packages are in the RSICC collection, and the majority are applicable to reactor dosimetry.

  16. The Radiation Safety Information Computational Center:. a Resource for Reactor Dosimetry Software and Nuclear Data

    NASA Astrophysics Data System (ADS)

    Kirk, B. L.

    2009-08-01

    The Radiation Safety Information Computational Center (RSICC) was established in 1963 to collect and disseminate computational nuclear technology in the form of radiation transport, shielding and safety software and corresponding nuclear cross sections. Approximately 1700 nuclear software and data packages are in the RSICC collection, and the majority are applicable to reactor dosimetry.

  17. 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 PLANTS AND MATERIALS Physical Protection Requirements at Fixed Sites § 73.58 Safety/security...

  18. 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 PLANTS AND MATERIALS Physical Protection Requirements at Fixed Sites § 73.58 Safety/security...

  19. 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 PLANTS AND MATERIALS Physical Protection Requirements at Fixed Sites § 73.58 Safety/security...

  20. Safety system augmentation at Russian nuclear power plants

    SciTech Connect

    Scerbo, J.A.; Satpute, S.N.; Donkin, J.Y.; Reister, R.A. |

    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 power 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.

  1. Safety aspect concerning radiolytic gas generation in reactors.

    PubMed

    Ramshesh, V

    2001-01-01

    In water cooled and water moderated reactors (H2O in boiling water reactors/pressurised water reactors, D2O in pressurised heavy water reactors) during normal operation, radiolysis is a source of production of hydrogen/deuterium and oxygen. During the progress of a nuclear accident, while there are other important sources of hydrogen/deuterium, the oxygen availability can occur only through radiolysis or direct contact with air. In air saturated with water vapour at room temperature and pressure when H2/D2 concentration exceeds 4 vol % (a conservative estimate), a combustible mixture with oxygen can be formed. It is proposed to examine the basic principles of water radiolysis as far as they pertain to generation of H2/D2 and O2 and try to apply these concepts to reactors both under operating conditions and in accident situations. It is concluded that the possibility of an accident taking place through radiolysis is highly unlikely. PMID:11382138

  2. Radionuclide inventories for short run-time space nuclear reactor systems

    NASA Astrophysics Data System (ADS)

    Coats, Richard L.

    1993-01-01

    Space Nuclear Reactor Systems, especially those used for propulsion, often have expected operation run times much shorter than those for land-based nuclear power plants. This produces substantially different radionuclide inventories to be considered in the safety analyses of space nuclear systems. This presentation describes an analysis utilizing ORIGEN2 and DKPOWER to provide comparisons among representative land-based and space systems. These comparisons enable early, conceptual considerations of safety issues and features in the preliminary design phases of operational systems, test facilities, and operations by identifying differences between the requirements for space systems and the established practice for land-based power systems. Early indications are that separation distance is much more effective as a safety measure for space nuclear systems than for power reactors because greater decay of the radionuclide activity occurs during the time to transport the inventory a given distance. In addition, the inventories of long-lived actinides are very low for space reactor systems.

  3. Radionuclide inventories for short run-time space nuclear reactor systems

    NASA Astrophysics Data System (ADS)

    Coats, R. L.

    1992-10-01

    Space Nuclear Reactor Systems, especially those used for propulsion, often have expected operation run times much shorter than those for land-based nuclear power plants. This produces substantially different radionuclide inventories to be considered in the safety analyses of space nuclear systems. This presentation describes an analysis utilizing ORIGEN2 and DKPOWER to provide comparisons among representative land-based and space systems. These comparisons enable early, conceptual considerations of safety issues and features in the preliminary design phases of operational systems, test facilities, and operations by identifying differences between the requirements for space systems and the established practice for land-based power systems. Early indications are that separation distance is much more effective as a safety measure for space nuclear systems than for power reactors because greater decay of the radionuclide activity occurs during the time to transport the inventory a given distance. In addition, the inventories of long-lived actinides are very low for space reactor systems.

  4. Radiological Safety Analysis Code System.

    Energy Science and Technology Software Center (ESTSC)

    2009-12-22

    Version 03 RSAC-6.2 can be used to model complex accidents and radiological consequences to individuals from the release of radionuclides to the atmosphere. A user can generate a fission product inventory; decay and ingrow the inventory during transport through processes, facilities, and the environment; model the downwind dispersion of the activity; and calculate doses to downwind individuals. Doses are calculated through the inhalation, immersion, ground surface and ingestion pathways. New to RSAC-6.2 are the abilitiesmore » to calculate inhalation from release to a room, inhalation from resuspension of activities, and a new model for dry deposition. Doses can now be calculated as close as 10 meters from the release point. RSAC-6.2 has been subjected to extensive independent verification and validation for use in performing safety-related dose calculations to support safety analysis reports. WinRP 2.0, a windows based overlay to RSAC-6.2, assists users in creating and running RSAC-6.2 input files. RSAC-6, Rev. 6.2 (03/11/02) corrects an earlier issue with RSAC-6, compiled with F77L-EM/32 Fortran 77 Version 5.10, which would not allow the executable to run with XP or VISTA Windows operating systems. Because this version is still in use at some facilities, it is being released through RSICC in addition to the new RSAC 7 (CCC-761).« less

  5. INTEGRATED SAFETY MANAGEMENT SYSTEM SAFETY CULTURE IMPROVEMENT INITIATIVE

    SciTech Connect

    MCDONALD JA JR

    2009-01-16

    In 2007, the Department of Energy (DOE) identified safety culture as one of their top Integrated Safety Management System (ISMS) related priorities. A team was formed to address this issue. The team identified a consensus set of safety culture principles, along with implementation practices that could be used by DOE, NNSA, and their contractors. Documented improvement tools were identified and communicated to contractors participating in a year long pilot project. After a year, lessons learned will be collected and a path forward determined. The goal of this effort was to achieve improved safety and mission performance through ISMS continuous improvement. The focus of ISMS improvement was safety culture improvement building on operating experience from similar industries such as the domestic and international commercial nuclear and chemical industry.

  6. 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. PMID:21399407

  7. The Oak Ridge Research Reactor: Safety analysis: Volume 2, Supplement 3

    SciTech Connect

    Cook, D.H.; Hamrick, T.P.

    1987-06-29

    The Oak Ridge Research Reactor (ORR) was constructed in the mid 1950s. Since it is an older facility, the issue of life-limiting conditions or material deterioration resulting from prolonged exposure to the normal operating environment is an item that should be addressed in the safety analysis for the ORR. Life-limiting conditions were considered in the original design of ORR; but due to the limited data that were available at that time on material performance in research reactors, various studies were completed during the first 10 years of operation at ORR to verify the applicable life-limiting parameters. Based on today's knowledge of life limiting conditions and the previous 30 years of operating experience at the ORR facility, the three specific areas of concern are addressed in this supplement: (1) embrittlement of the structures due to radiation damage, which is described in Section 2; (2) fatigue due to the effects of both thermal cycling and vibration, which is addressed in Section 3; and (3) the effects of corrosion on the integrity of the primary system, which is described in Section 4. The purpose of this document is to provide a review of the applicable safety studies which have been performed, and to state the status of the ORR with regard to embrittlement, fatigue (due to thermal cycling and vibration), and corrosion.

  8. Developing a Comprehensive Software Suite for Advanced Reactor Performance and Safety Analysis

    SciTech Connect

    Pointer, William David; Bradley, Keith S; Fischer, Paul F; Smith, Micheal A; Tautges, Timothy J; Ferencz, Robert M; Martineau, Richard C; Jain, Rajeev; Obabko, Aleksandr; Billings, Jay Jay

    2013-01-01

    This paper provides an introduction to the reactor analysis capabilities of the nuclear power reactor simulation tools that are being developed as part of the U.S. Department of Energy s Nuclear Energy Advanced Modeling and Simulation (NEAMS) Toolkit. The NEAMS Toolkit is an integrated suite of multi-physics simulation tools that leverage high-performance computing to reduce uncertainty in the prediction of performance and safety of advanced reactor and fuel designs. The Toolkit effort is comprised of two major components, the Fuels Product Line (FPL), which provides tools for fuel performance analysis, and the Reactor Product Line (RPL), which provides tools for reactor performance and safety analysis. This paper provides an overview of the NEAMS RPL development effort.

  9. Guidelines for preparing criticality safety evaluations at Department of Energy non-reactor nuclear facilities

    SciTech Connect

    1998-09-01

    This Department of Energy (DOE) is approved for use by all components of DOE. It contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE Non-Reactor Nuclear Facilities. Adherence with these guidelines will provide consistency and uniformity in Criticality Safety Evaluations (CSEs) across the complex and will document compliance with DOE Order 5480.24 requirements as they pertain to CSEs.

  10. Software Quality Assurance for Nuclear Safety Systems

    SciTech Connect

    Sparkman, D R; Lagdon, R

    2004-05-16

    The US Department of Energy has undertaken an initiative to improve the quality of software used to design and operate their nuclear facilities across the United States. One aspect of this initiative is to revise or create new directives and guides associated with quality practices for the safety software in its nuclear facilities. Safety software includes the safety structures, systems, and components software and firmware, support software and design and analysis software used to ensure the safety of the facility. DOE nuclear facilities are unique when compared to commercial nuclear or other industrial activities in terms of the types and quantities of hazards that must be controlled to protect workers, public and the environment. Because of these differences, DOE must develop an approach to software quality assurance that ensures appropriate risk mitigation by developing a framework of requirements that accomplishes the following goals: {sm_bullet} Ensures the software processes developed to address nuclear safety in design, operation, construction and maintenance of its facilities are safe {sm_bullet} Considers the larger system that uses the software and its impacts {sm_bullet} Ensures that the software failures do not create unsafe conditions Software designers for nuclear systems and processes must reduce risks in software applications by incorporating processes that recognize, detect, and mitigate software failure in safety related systems. It must also ensure that fail safe modes and component testing are incorporated into software design. For nuclear facilities, the consideration of risk is not necessarily sufficient to ensure safety. Systematic evaluation, independent verification and system safety analysis must be considered for software design, implementation, and operation. The software industry primarily uses risk analysis to determine the appropriate level of rigor applied to software practices. This risk-based approach distinguishes safety

  11. Lessons Learned in the Update of a Safety Limit for the High Flux Isotope Reactor

    SciTech Connect

    Cook, David Howard

    2009-01-01

    A recent unreviewed safety question (USQ) regarding a portion of the High Flux Isotope Reactor (HFIR) transient decay heat removal analysis focused on applicability of a heat transfer correlation at the low flow end of reactor operations. During resolution of this issue, review of the correlations used to establish the safety limit (SL) on reactor flux-to-flow ratio revealed the need to change the magnitude of the SL at the low flow end of reactor operations and the need to update the hot spot fuel damage criteria to incorporate current knowledge involving parallel channel flow stability. Because of the original safety design strategy for the reactor, resolution of the issues for the flux-to-flow ratio involved reevaluation of all key process variable SLs and limiting control settings (LCSs) using the current version of the heat transfer analysis code for the reactor. Goals of the work involved updating and upgrading the SL analysis where necessary, while preserving the safety design strategy for the reactor. Changes made include revisions to the safety design criteria at low flows to address the USQ, update of the process- and analysis input-variable uncertainty considerations, and upgrade of the safety design criteria at high flow. The challenges faced during update/upgrade of this SL and LCS are typical of the problems found in the integration of safety into the design process for a complex facility. In particular, the problems addressed in the area of instrument uncertainties provide valuable lessons learned for establishment and configuration control of SLs for large facilities.

  12. 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," focuses on a variety…

  13. Gaseous fuel reactors for power systems

    NASA Technical Reports Server (NTRS)

    Kendall, J. S.; Rodgers, R. J.

    1977-01-01

    Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

  14. 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

  15. An overview of modeling methods for thermal mixing and stratification in large enclosures for reactor safety analysis

    SciTech Connect

    Haihua Zhao; Per F. Peterson

    2010-10-01

    Thermal mixing and stratification phenomena play major roles in the safety of reactor systems with large enclosures, such as containment safety in current fleet of LWRs, long-term passive containment cooling in Gen III+ plants including AP-1000 and ESBWR, the cold and hot pool mixing in pool type sodium cooled fast reactor systems (SFR), and reactor cavity cooling system behavior in high temperature gas cooled reactors (HTGR), etc. Depending on the fidelity requirement and computational resources, 0-D steady state models (heat transfer correlations), 0-D lumped parameter based transient models, 1-D physical-based coarse grain models, and 3-D CFD models are available. Current major system analysis codes either have no models or only 0-D models for thermal stratification and mixing, which can only give highly approximate results for simple cases. While 3-D CFD methods can be used to analyze simple configurations, these methods require very fine grid resolution to resolve thin substructures such as jets and wall boundaries. Due to prohibitive computational expenses for long transients in very large volumes, 3-D CFD simulations remain impractical for system analyses. For mixing in stably stratified large enclosures, UC Berkeley developed 1-D models basing on Zuber’s hierarchical two-tiered scaling analysis (HTTSA) method where the ambient fluid volume is represented by 1-D transient partial differential equations and substructures such as free or wall jets are modeled with 1-D integral models. This allows very large reductions in computational effort compared to 3-D CFD modeling. This paper will present an overview on important thermal mixing and stratification phenomena in large enclosures for different reactors, major modeling methods and their advantages and limits, potential paths to improve simulation capability and reduce analysis uncertainty in this area for advanced reactor system analysis tools.

  16. Fast reactor safety program. Progress report, January-March 1980

    SciTech Connect

    1980-05-01

    The goal of the DOE LMFBR Safety Program is to provide a technology base fully responsive to safety considerations in the design, evaluation, licensing, and economic optimization of LMFBRs for electrical power generation. A strategy is presented that divides safety technology development into seven program elements, which have been used as the basis for the Work Breakdown Structure (WBS) for the Program. These elements include four lines of assurance (LOAs) involving core-related safety considerations, an element supporting non-core-related plant safety considerations, a safety R and D integration element, and an element for the development of test facilities and equipment to be used in Program experiments: LOA-1 (prevent accidents); LOA-2 (limit core damage); LOA-3 (maintain containment integrity); LOA-4 (attenuate radiological consequences); plant considerations; R and D integration; and facility development.

  17. Reactor with natural convection backup cooling system

    SciTech Connect

    Koutz, S.L.; Cavallaro, L.; Kapich, D.D.

    1988-07-26

    A nuclear reactor is described comprising: a closed vessel containing a pool of secondary coolant; a reactor core located within the pool; an inlet conduit system for directing primary coolant to the lower end of the core; pumping means for effecting flow of the primary coolant into the inlet conduit system; an outlet conduit system located above the core for receiving coolant from the upper end of the core and carrying it to an exterior outlet pipe; and pressure reduction means for reducing fluid pressure in the inlet conduit system adjacent all of the one or more opening therein by locally increasing fluid velocity adjacent all of the one or more openings.

  18. Control system for a small fission reactor

    DOEpatents

    Burelbach, James P.; Kann, William J.; Saiveau, James G.

    1986-01-01

    A system for controlling the reactivity of a small fission reactor includes an elongated, flexible hollow tube in the general form of a helical coiled spring axially positioned around and outside of the reactor vessel in an annular space between the reactor vessel and a surrounding cylindrical-shaped neutron reflector. A neutron absorbing material is provided within the hollow tube with the rate of the reaction controlled by the extension and compression of the hollow tube, e.g., extension of the tube increases reactivity while its compression reduces reactivity, in varying the amount of neutron absorbing material disposed between the reactor vessel and the neutron reflector. Conventional mechanical displacement means may be employed to control the coil density of the hollow tube as desired. In another embodiment, a plurality of flexible hollow tubes each containing a neutron absorber are positioned adjacent to one another in spaced relation around the periphery of the reactor vessel and inside the outer neutron reflector with reactivity controlled by the extension and compression of all or some of the coiled hollow tubes. Yet another embodiment of the invention envisions the neutron reflector in the form of an expandable coil spring positioned in an annular space between the reactor vessel and an outer neutron absorbing structure for controlling the neutron flux reflected back into the reactor vessel.

  19. A philosophy for space nuclear systems safety

    NASA Astrophysics Data System (ADS)

    Marshall, A. C.

    The unique requirements and contraints of space nuclear systems require careful consideration in the development of a safety policy. The Nuclear Safety Policy Working Group (NSPWG) for the Space Exploration Initiative has proposed a hierarchical approach with safety policy at the top of the hierarchy. This policy allows safety requirements to be tailored to specific applications while still providing reassurance to regulators and the general public that the necessary measures have been taken to assure safe application of space nuclear systems. The safety policy used by the NSPWG is recommended for all space nuclear programs and missions.

  20. A philosophy for space nuclear systems safety

    SciTech Connect

    Marshall, A.C.

    1992-08-01

    The unique requirements and contraints of space nuclear systems require careful consideration in the development of a safety policy. The Nuclear Safety Policy Working Group (NSPWG) for the Space Exploration Initiative has proposed a hierarchical approach with safety policy at the top of the hierarchy. This policy allows safety requirements to be tailored to specific applications while still providing reassurance to regulators and the general public that the necessary measures have been taken to assure safe application of space nuclear systems. The safety policy used by the NSPWG is recommended for all space nuclear programs and missions.

  1. Code System to Calculate Reactor Coolant System Leak Rate.

    SciTech Connect

    Bell, Pat

    1999-10-19

    Version 00 RCSLK9 was developed to analyze the leak tightness of the primary coolant system for any pressurized water reactor (PWR). From given system conditions, water levels in tanks, and certain system design parameters, RCSLK9 calculates the loss of water from the reactor coolant system (RCS) and the increase of water in the leakage collection system during an arbitrary time interval. The program determines the system leak rates and displays or prints a report of the results. During the initial application to a specific reactor, RCSLK9 creates a file of system parameters and saves it for future use.

  2. Code System to Calculate Reactor Coolant System Leak Rate.

    Energy Science and Technology Software Center (ESTSC)

    1999-10-19

    Version 00 RCSLK9 was developed to analyze the leak tightness of the primary coolant system for any pressurized water reactor (PWR). From given system conditions, water levels in tanks, and certain system design parameters, RCSLK9 calculates the loss of water from the reactor coolant system (RCS) and the increase of water in the leakage collection system during an arbitrary time interval. The program determines the system leak rates and displays or prints a report ofmore » the results. During the initial application to a specific reactor, RCSLK9 creates a file of system parameters and saves it for future use.« less

  3. Professional Issues in System Safety Engineering

    NASA Astrophysics Data System (ADS)

    McDermid, John; Thomas, Martyn; Redmill, Felix

    For many years the profession of system safety engineering has been emerging. This paper argues that the time has now come when it requires recognition, a voice, proper governance and leadership. System safety engineering is an amalgam of many disciplines, in particular, software engineering, safety engineering and management, and systems engineering, and this paper shows that system safety engineering must address the most difficult aspects of all of these. But professional matters extend beyond merely technical considerations, and the paper concludes by showing why there is the need for a new professional body.

  4. Improved methodology for integral analysis of advanced reactors employing passive safety

    NASA Astrophysics Data System (ADS)

    Muftuoglu, A. Kursad

    After four decades of experience with pressurized water reactors, a new generation of nuclear plants are emerging. These advanced designs employ passive safety which relies on natural forces, such as gravity and natural circulation. The new concept of passive safety also necessitates improvement in computational tools available for best-estimate analyses. The system codes originally designed for high pressure conditions in the presence of strong momentum sources such as pumps are challenged in many ways. Increased interaction of the primary system with the containment necessitates a tool for integral analysis. This study addresses some of these concerns. An improved tool for integral analysis coupling primary system with containment calculation is also presented. The code package is based on RELAP5 and CONTAIN programs, best-estimate thermal-hydraulics code for primary system analysis and containment code for containment analysis, respectively. The suitability is demonstrated with a postulated small break loss of coolant accident analysis of Westinghouse AP600 plant. The thesis explains the details of the analysis including the coupling model.

  5. Automation for System Safety Analysis

    NASA Technical Reports Server (NTRS)

    Malin, Jane T.; Fleming, Land; Throop, David; Thronesbery, Carroll; Flores, Joshua; Bennett, Ted; Wennberg, Paul

    2009-01-01

    This presentation describes work to integrate a set of tools to support early model-based analysis of failures and hazards due to system-software interactions. The tools perform and assist analysts in the following tasks: 1) extract model parts from text for architecture and safety/hazard models; 2) combine the parts with library information to develop the models for visualization and analysis; 3) perform graph analysis and simulation to identify and evaluate possible paths from hazard sources to vulnerable entities and functions, in nominal and anomalous system-software configurations and scenarios; and 4) identify resulting candidate scenarios for software integration testing. There has been significant technical progress in model extraction from Orion program text sources, architecture model derivation (components and connections) and documentation of extraction sources. Models have been derived from Internal Interface Requirements Documents (IIRDs) and FMEA documents. Linguistic text processing is used to extract model parts and relationships, and the Aerospace Ontology also aids automated model development from the extracted information. Visualizations of these models assist analysts in requirements overview and in checking consistency and completeness.

  6. Advanced reactor safety program. Stakeholder interaction and feedback

    SciTech Connect

    Szilard, Ronaldo H.; Smith, Curtis L.

    2014-08-01

    In the Spring of 2013, we began discussions with our industry stakeholders on how to upgrade our safety analysis capabilities. The focus of these improvements would primarily be on advanced safety analysis capabilities that could help the nuclear industry analyze, understand, and better predict complex safety problems. The current environment in the DOE complex is such that recent successes in high performance computer modeling could lead the nuclear industry to benefit from these advances, as long as an effort to translate these advances into realistic applications is made. Upgrading the nuclear industry modeling analysis capabilities is a significant effort that would require substantial participation and coordination from all industry segments: research, engineering, vendors, and operations. We focus here on interactions with industry stakeholders to develop sound advanced safety analysis applications propositions that could have a positive impact on industry long term operation, hence advancing the state of nuclear safety.

  7. Control system for a small fission reactor

    DOEpatents

    Burelbach, J.P.; Kann, W.J.; Saiveau, J.G.

    1985-02-08

    A system for controlling the reactivity of a small fission reactor includes an elongated, flexible hollow tube in the general form of a helical coiled spring axially positioned around and outside of the reactor vessel in an annular space between the reactor vessel and a surrounding cylindrical-shaped neutron reflector. A neutron absorbing material is provided within the hollow tube with the rate of the reaction controlled by the extension and compression of the hollow tube, e.g., extension of the tube increases reactivity while its compression reduces reactivity, in varying the amount of neutron absorbing material disposed between the reactor vessel and the neutron reflector. Conventional mechanical displacement means may be employed to control the coil density of the hollow tube as desired.

  8. NEUTRONIC REACTOR FUEL ELEMENT AND CORE SYSTEM

    DOEpatents

    Moore, W.T.

    1958-09-01

    This patent relates to neutronic reactors and in particular to an improved fuel element and a novel reactor core system for facilitating removal of contaminating fission products, as they are fermed, from association with the flssionable fuel, so as to mitigate the interferent effects of such fission products during reactor operation. The fuel elements are comprised of tubular members impervious to fluid and contatning on their interior surfaces a thin layer of fissionable material providing a central void. The core structure is comprised of a plurality of the tubular fuel elements arranged in parallel and a closed manifold connected to their ends. In the reactor the core structure is dispersed in a water moderator and coolant within a pressure vessel, and a means connected to said manifuld is provided for withdrawing and disposing of mobile fission product contamination from the interior of the feel tubes and manifold.

  9. Software safety and reliability issues in safety-related systems

    SciTech Connect

    Zucconi, L.

    1992-09-01

    The increasing number of accidents attributed to computer-based systems is causing increased public awareness of the risk associated with these systems' use in safety-related applications. Examples include the Therac-25 medical LINAC deaths, the growing number of Airbus A320 crashes, the AT T Long-Lines disaster on Martin Luther King Day in 1990, the spate of regional telephone outages of the summer of 1991, and many more. How do safety and reliability sometimes conflict What practical computer system and software development technologies and processes can be applied to increase the safety and reliability of computer systems What are the technical and managerial issues contributing to the construction of less-than-safe computer-based systems How can systems engineers and software engineers work together. to address the issues related to safety and reliability of computer systems This paper will address these topics and include an assessment of the best current state-of-the-practice and upcoming technologies that will carry us into the 21st century.

  10. Safety and reliability issues in safety-related systems

    SciTech Connect

    Zucconi, L.

    1992-03-20

    The increasing number of accidents attributed to computer-based systems is causing increased public awareness of the risk associated with these systems` use in safety-related applications. Examples include the Therac-25 medical LINAC deaths, the growing number of Airbus A320 crashes, the AT&T Long-Lines disaster on Martin Luther King Day in 1990, the spate of regional telephone outages of the summer of 1991, and many more. How do safety and reliability sometimes conflict? What practical computer system and software development technologies and processes can be applied to increase the safety and reliability of computer systems? What are the technical and managerial issues contributing to the construction of less-than-safe computer-based systems? How can systems engineers and software engineers work together to address the issues related safety and reliability of computer systems? This paper will address these topics and include an assessment of the best current state-of-the-practice and of upcoming technologies that will carry us into the 21st century.

  11. Software safety and reliability issues in safety-related systems

    SciTech Connect

    Zucconi, L.

    1992-09-01

    The increasing number of accidents attributed to computer-based systems is causing increased public awareness of the risk associated with these systems` use in safety-related applications. Examples include the Therac-25 medical LINAC deaths, the growing number of Airbus A320 crashes, the AT&T Long-Lines disaster on Martin Luther King Day in 1990, the spate of regional telephone outages of the summer of 1991, and many more. How do safety and reliability sometimes conflict? What practical computer system and software development technologies and processes can be applied to increase the safety and reliability of computer systems? What are the technical and managerial issues contributing to the construction of less-than-safe computer-based systems? How can systems engineers and software engineers work together. to address the issues related to safety and reliability of computer systems? This paper will address these topics and include an assessment of the best current state-of-the-practice and upcoming technologies that will carry us into the 21st century.

  12. Safety and reliability issues in safety-related systems

    SciTech Connect

    Zucconi, L.

    1992-03-20

    The increasing number of accidents attributed to computer-based systems is causing increased public awareness of the risk associated with these systems' use in safety-related applications. Examples include the Therac-25 medical LINAC deaths, the growing number of Airbus A320 crashes, the AT T Long-Lines disaster on Martin Luther King Day in 1990, the spate of regional telephone outages of the summer of 1991, and many more. How do safety and reliability sometimes conflict What practical computer system and software development technologies and processes can be applied to increase the safety and reliability of computer systems What are the technical and managerial issues contributing to the construction of less-than-safe computer-based systems How can systems engineers and software engineers work together to address the issues related safety and reliability of computer systems This paper will address these topics and include an assessment of the best current state-of-the-practice and of upcoming technologies that will carry us into the 21st century.

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

    SciTech Connect

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

    2012-08-01

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

  14. DEVELOPMENT OF RISK-BASED AND TECHNOLOGY-INDEPENDENT SAFETY CRITERIA FOR GENERATION IV SYSTEMS

    SciTech Connect

    William E. Kastenberg; Edward Blandford; Lance Kim

    2009-03-31

    This project has developed quantitative safety goals for Generation IV (Gen IV) nuclear energy systems. These safety goals are risk based and technology independent. The foundations for a new approach to risk analysis has been developed, along with a new operational definition of risk. This project has furthered the current state-of-the-art by developing quantitative safety goals for both Gen IV reactors and for the overall Gen IV nuclear fuel cycle. The risk analysis approach developed will quantify performance measures, characterize uncertainty, and address a more comprehensive view of safety as it relates to the overall system. Appropriate safety criteria are necessary to manage risk in a prudent and cost-effective manner. This study is also important for government agencies responsible for managing, reviewing, and for approving advanced reactor systems because they are charged with assuring the health and safety of the public.

  15. The Evolution of System Safety at NASA

    NASA Technical Reports Server (NTRS)

    Dezfuli, Homayoon; Everett, Chris; Groen, Frank

    2014-01-01

    The NASA system safety framework is in the process of change, motivated by the desire to promote an objectives-driven approach to system safety that explicitly focuses system safety efforts on system-level safety performance, and serves to unify, in a purposeful manner, safety-related activities that otherwise might be done in a way that results in gaps, redundancies, or unnecessary work. An objectives-driven approach to system safety affords more flexibility to determine, on a system-specific basis, the means by which adequate safety is achieved and verified. Such flexibility and efficiency is becoming increasingly important in the face of evolving engineering modalities and acquisition models, where, for example, NASA will increasingly rely on commercial providers for transportation services to low-earth orbit. A key element of this objectives-driven approach is the use of the risk-informed safety case (RISC): a structured argument, supported by a body of evidence, that provides a compelling, comprehensible and valid case that a system is or will be adequately safe for a given application in a given environment. The RISC addresses each of the objectives defined for the system, providing a rational basis for making informed risk acceptance decisions at relevant decision points in the system life cycle.

  16. Processes and Procedures for Application of CFD to Nuclear Reactor Safety Analysis

    SciTech Connect

    Richard W. Johnson; Richard R. Schultz; Patrick J. Roache; Ismail B. Celik; William D. Pointer; Yassin A. Hassan

    2006-09-01

    Traditionally, nuclear reactor safety analysis has been performed using systems analysis codes such as RELAP5, which was developed at the INL. However, goals established by the Generation IV program, especially the desire to increase efficiency, has lead to an increase in operating temperatures for the reactors. This increase pushes reactor materials to operate towards their upper temperature limits relative to structural integrity. Because there will be some finite variation of the power density in the reactor core, there will be a potential for local hot spots to occur in the reactor vessel. Hence, it has become apparent that detailed analysis will be required to ensure that local ‘hot spots’ do not exceed safety limits. It is generally accepted that computational fluid dynamics (CFD) codes are intrinsically capable of simulating fluid dynamics and heat transport locally because they are based on ‘first principles.’ Indeed, CFD analysis has reached a fairly mature level of development, including the commercial level. However, CFD experts are aware that even though commercial codes are capable of simulating local fluid and thermal physics, great care must be taken in their application to avoid errors caused by such things as inappropriate grid meshing, low-order discretization schemes, lack of iterative convergence and inaccurate time-stepping. Just as important is the choice of a turbulence model for turbulent flow simulation. Turbulence models model the effects of turbulent transport of mass, momentum and energy, but are not necessarily applicable for wide ranges of flow types. Therefore, there is a well-recognized need to establish practices and procedures for the proper application of CFD to simulate flow physics accurately and establish the level of uncertainty of such computations. The present document represents contributions of CFD experts on what the basic practices, procedures and guidelines should be to aid CFD analysts to obtain accurate

  17. 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.

  18. Evaluation of proposed German safety criteria for high-temperature gas-cooled reactors

    SciTech Connect

    Barsell, A.W.

    1980-05-01

    This work reviews proposed safety criteria prepared by the German Bundesministerium des Innern (BMI) for future licensing of gas-cooled high-temperature reactor (HTR) concepts in the Federal Republic of Germany. Comparison is made with US General Design Criteria (GDCs) in 10CFR50 Appendix A and with German light water reactor (LWR) criteria. Implications for the HTR design relative to the US design and safety approach are indicated. Both inherent characteristics and design features of the steam cycle, gas turbine, and process heat concepts are taken into account as well as generic design options such as a pebble bed or prismatic core.

  19. Neutronic safety parameters and transient analyses for potential LEU conversion of the Budapest Research Reactor.

    SciTech Connect

    Pond, R. B.; Hanan, N. A.; Matos, J. E.; Maraczy, C.

    1999-09-27

    An initial safety study for potential LEU conversion of the Budapest Research Reactor was completed. The study compares safety parameters and example transients for reactor cores with HEU and LEU fuels. Reactivity coefficients, kinetic parameters and control rod worths were calculated for cores with HEU(36%) UAl alloy fuel and UO{sub 2}-Al dispersion fuel, and with LEU (19.75%)UO{sub 2}-Al dispersion fuel that has a uranium density of about 2.5 g/cm{sup 3}. A preliminary fuel conversion plan was developed for transition cores that would convert the BRR from HEU to LEU fuel after the process is begun.

  20. Computer analyses for the design, operation and safety of new isotope production reactors: A technology status review

    SciTech Connect

    Wulff, W.

    1990-01-01

    A review is presented on the currently available technologies for nuclear reactor analyses by computer. The important distinction is made between traditional computer calculation and advanced computer simulation. Simulation needs are defined to support the design, operation, maintenance and safety of isotope production reactors. Existing methods of computer analyses are categorized in accordance with the type of computer involved in their execution: micro, mini, mainframe and supercomputers. Both general and special-purpose computers are discussed. Major computer codes are described, with regard for their use in analyzing isotope production reactors. It has been determined in this review that conventional systems codes (TRAC, RELAP5, RETRAN, etc.) cannot meet four essential conditions for viable reactor simulation: simulation fidelity, on-line interactive operation with convenient graphics, high simulation speed, and at low cost. These conditions can be met by special-purpose computers (such as the AD100 of ADI), which are specifically designed for high-speed simulation of complex systems. The greatest shortcoming of existing systems codes (TRAC, RELAP5) is their mismatch between very high computational efforts and low simulation fidelity. The drift flux formulation (HIPA) is the viable alternative to the complicated two-fluid model. No existing computer code has the capability of accommodating all important processes in the core geometry of isotope production reactors. Experiments are needed (heat transfer measurements) to provide necessary correlations. It is important for the nuclear community, both in government, industry and universities, to begin to take advantage of modern simulation technologies and equipment. 41 refs.

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

    SciTech Connect

    Monteleone, S.

    1993-10-01

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

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

    SciTech Connect

    Monteleone, S.

    1997-09-01

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

  3. Dynamic Impregnator Reactor System (Poster)

    SciTech Connect

    Not Available

    2012-09-01

    IBRF poster developed for the IBRF showcase. Describes the multifarious system designed for complex feedstock impregnation and processing. IBRF feedstock system has several unit operations combined into one robust system that provides for flexible and staged process configurations, such as spraying, soaking, low-severity pretreatment, enzymatic hydrolysis, fermentation, concentration/evaporation, and distillation.

  4. On the Criticality Safety of Transuranic Sodium Fast Reactor Fuel Transport Casks

    SciTech Connect

    Samuel Bays; Ayodeji Alajo

    2010-05-01

    This work addresses the neutronic performance and criticality safety issues of transport casks for fuel pertaining to low conversion ratio sodium cooled fast reactors, conventionally known as Advanced Burner Reactors. The criticality of a one, three, seven and 19-assembly cask capacity is presented. Both dry “helium” and flooded “water” filled casks are considered. No credit for fuel burnup or fission products was assumed. As many as possible of the conservatisms used in licensing light water reactor universal transport casks were incorporated into this SFR cask criticality design and analysis. It was found that at 7-assemblies or more, adding moderator to the SFR cask increases criticality margin. Also, removal of MAs from the fuel increases criticality margin of dry casks and takes a slight amount of margin away for wet casks. Assuming credit for borated fuel tube liners, this design analysis suggests that as many as 19 assemblies can be loaded in a cask if limited purely by criticality safety. If no credit for boron is assumed, the cask could possibly hold seven assemblies if low conversion ratio fast reactor grade fuel and not breeder reactor grade fuel is assumed. The analysis showed that there is a need for new cask designs for fast reactors spent fuel transportation. There is a potential of modifying existing transportation cask design as the starting point for fast reactor spent fuel transportation.

  5. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Main report

    SciTech Connect

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). The study results are presented in two volumes. Volume 1 (Main Report) contains the results in summary form.

  6. The uses and benefits of probabilistic risk assessment in nuclear reactor safety

    SciTech Connect

    Bari, R.A.; Speis, T.P.; Nuclear Regulatory Commission, Washington, DC . Office of Nuclear Regulatory Research)

    1989-01-01

    Probabilistic risk assessment (PRA) has proven to be an important tool in the safety assessment of nuclear reactors throughout the world. Decision making with regard to many safety issues has been facilitated by both general insights from and direct application of this technology. Key uses of PRA are discussed and some examples of successful applications are cited. The benefits and limitations of PRA are also discussed as well as the broader outlook for applications of PRA. 9 refs.

  7. Safety goals and functional performance criteria. [Advanced Reactor Design

    SciTech Connect

    Youngblood, R.W.; Pratt, W.T. ); Hardin, W.B. . Div. of Regulatory Applications)

    1990-01-01

    This report discusses a possible approach to the development of functional performance criteria to be applied to evolutionary LWR designs. Key safety functions are first identified; then, criteria are drawn up for each individual function, based on the premise that no single function's projected unreliability should be allowed to exhaust the safety goal frequencies. In the area of core damage prevention, functional criteria are cast in terms of necessary levels of redundancy and diversity of critical equipment. In the area of core damage mitigation (containment), functional performance criteria are cast with the aim of mitigating post-core-melt phenomena with sufficient assurance to eliminate major uncertainties in containment performance. 9 refs.

  8. Exploration of high-dimensional scalar function for nuclear reactor safety analysis and visualization

    SciTech Connect

    Maljovec, D.; Wang, B.; Pascucci, V.; Bremer, P. T.; Pernice, M.; Mandelli, D.; Nourgaliev, R.

    2013-07-01

    The next generation of methodologies for nuclear reactor Probabilistic Risk Assessment (PRA) explicitly accounts for the time element in modeling the probabilistic system evolution and uses numerical simulation tools to account for possible dependencies between failure events. The Monte-Carlo (MC) and the Dynamic Event Tree (DET) approaches belong to this new class of dynamic PRA methodologies. A challenge of dynamic PRA algorithms is the large amount of data they produce which may be difficult to visualize and analyze in order to extract useful information. We present a software tool that is designed to address these goals. We model a large-scale nuclear simulation dataset as a high-dimensional scalar function defined over a discrete sample of the domain. First, we provide structural analysis of such a function at multiple scales and provide insight into the relationship between the input parameters and the output. Second, we enable exploratory analysis for users, where we help the users to differentiate features from noise through multi-scale analysis on an interactive platform, based on domain knowledge and data characterization. Our analysis is performed by exploiting the topological and geometric properties of the domain, building statistical models based on its topological segmentations and providing interactive visual interfaces to facilitate such explorations. We provide a user's guide to our software tool by highlighting its analysis and visualization capabilities, along with a use case involving data from a nuclear reactor safety simulation. (authors)

  9. Exploration of High-Dimensional Scalar Function for Nuclear Reactor Safety Analysis and Visualization

    SciTech Connect

    Dan Maljovec; Bei Wang; Valerio Pascucci; Peer-Timo Bremer; Michael Pernice; Robert Nourgaliev

    2013-05-01

    The next generation of methodologies for nuclear reactor Probabilistic Risk Assessment (PRA) explicitly accounts for the time element in modeling the probabilistic system evolution and uses numerical simulation tools to account for possible dependencies between failure events. The Monte-Carlo (MC) and the Dynamic Event Tree (DET) approaches belong to this new class of dynamic PRA methodologies. A challenge of dynamic PRA algorithms is the large amount of data they produce which may be difficult to visualize and analyze in order to extract useful information. We present a software tool that is designed to address these goals. We model a large-scale nuclear simulation dataset as a high-dimensional scalar function defined over a discrete sample of the domain. First, we provide structural analysis of such a function at multiple scales and provide insight into the relationship between the input parameters and the output. Second, we enable exploratory analysis for users, where we help the users to differentiate features from noise through multi-scale analysis on an interactive platform, based on domain knowledge and data characterization. Our analysis is performed by exploiting the topological and geometric properties of the domain, building statistical models based on its topological segmentations and providing interactive visual interfaces to facilitate such explorations. We provide a user’s guide to our software tool by highlighting its analysis and visualization capabilities, along with a use case involving dataset from a nuclear reactor safety simulation.

  10. Hybrid Molten Salt Reactor (HMSR) System Study

    SciTech Connect

    Woolley, Robert D; Miller, Laurence F

    2014-04-01

    Can the hybrid system combination of (1) a critical fission Molten Salt Reactor (MSR) having a thermal spectrum and a high Conversion Ratio (CR) with (2) an external source of high energy neutrons provide an attractive solution to the world's expanding demand for energy? The present study indicates the answer is an emphatic yes.

  11. Microchannel Reactor System for Catalytic Hydrogenation

    SciTech Connect

    2004-07-01

    This factsheet describes a research project whose goal is to design, fabricate, evaluate, and optimize a laboratory-scale microchannel reactor/heat exchanger system with thin-film or particulate catalysts for hydrogenation of o-nitroanisole and other nitro aromatic compounds, under moderate temperature and pressure.

  12. NON-CORROSIVE REACTOR FUEL SYSTEM

    DOEpatents

    Herrick, C.C.

    1962-08-14

    A non-corrosive nuclear reactor fuel system was developed utilizing a molten plutonium-- iron alloy fuel having about 2 at.% carbon and contained in a tantalum vessel. This carbon reacts with the interior surface of the tantalum vessel to form a plutonium resistant self-healing tantalum carbide film. (AEC)

  13. System safety management: A new discipline

    NASA Technical Reports Server (NTRS)

    Pope, W. C.

    1971-01-01

    The systems theory is discussed in relation to safety management. It is suggested that systems safety management, as a new discipline, holds great promise for reducing operating errors, conserving labor resources, avoiding operating costs due to mistakes, and for improving managerial techniques. It is pointed out that managerial failures or system breakdowns are the basic reasons for human errors and condition defects. In this respect, a recommendation is made that safety engineers stop visualizing the problem only with the individual (supervisor or employee) and see the problem from the systems point of view.

  14. The CANDU Reactor System: An Appropriate Technology.

    PubMed

    Robertson, J A

    1978-02-10

    CANDU power reactors are characterized by the combination of heavy water as moderator and pressure tubes to contain the fuel and coolant. Their excellent neutron economy provides the simplicity and low costs of once-through natural-uranium fueling. Future benefits include the prospect of a near-breeder thorium fuel cycle to provide security of fuel supply without the need to develop a new reactor such as the fast breeder. These and other features make the CANDU system an appropriate technology for countries, like Canada, of intermediate economic and industrial capacity. PMID:17788102

  15. Rodded shutdown system for a nuclear reactor

    DOEpatents

    Golden, Martin P.; Govi, Aldo R.

    1978-01-01

    A top mounted nuclear reactor diverse rodded shutdown system utilizing gas fed into a pressure bearing bellows region sealed at the upper extremity to an armature. The armature is attached to a neutron absorber assembly by a series of shafts and connecting means. The armature is held in an uppermost position by an electromagnet assembly or by pressurized gas in a second embodiment. Deenergizing the electromagnet assembly, or venting the pressurized gas, causes the armature to fall by the force of gravity, thereby lowering the attached absorber assembly into the reactor core.

  16. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 1 2014-01-01 2014-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems...

  17. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 1 2012-01-01 2012-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems...

  18. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 1 2013-01-01 2013-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems...

  19. Gaseous fuel reactor systems for aerospace applications

    NASA Technical Reports Server (NTRS)

    Thom, K.; Schwenk, F. C.

    1977-01-01

    Research on the gaseous fuel nuclear rocket concept continues under the programs of the U.S. National Aeronautics and Space Administration (NASA) Office for Aeronautics and Space Technology and now includes work related to power applications in space and on earth. In a cavity reactor test series, initial experiments confirmed the low critical mass determined from reactor physics calculations. Recent work with flowing UF6 fuel indicates stable operation at increased power levels. Preliminary design and experimental verification of test hardware for high-temperature experiments have been accomplished. Research on energy extraction from fissioning gases has resulted in lasers energized by fission fragments. Combined experimental results and studies indicate that gaseous-fuel reactor systems have significant potential for providing nuclear fission power in space and on earth.

  20. Gaseous fuel reactors for power systems

    NASA Technical Reports Server (NTRS)

    Helmick, H. H.; Schwenk, F. C.

    1978-01-01

    The Los Alamos Scientific Laboratory is participating in a NASA-sponsored program to demonstrate the feasibility of a gaseous uranium fueled reactor. The work is aimed at acquiring experimental and theoretical information for the design of a prototype plasma core reactor which will test heat removal by optical radiation. The basic goal of this work is for space applications, however, other NASA-sponsored work suggests several attractive applications to help meet earth-bound energy needs. Such potential benefits are: small critical mass, on-site fuel processing, high fuel burnup, low fission fragment inventory in reactor core, high temperature for process heat, optical radiation for photochemistry and space power transmission, and high temperature for advanced propulsion systems.

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

    SciTech Connect

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

    2006-04-01

    the manufacturers site (Barber-Nichols Inc.) and installed and operated at Sandia. A sufficiently detailed description of the loop is provided in this report along with the design characteristics of the turbo-alternator-compressor set to allow other researchers to compare their results with those measured in the Sandia test-loop. The third task consisted of a validation effort. In this task the test loop was operated and compared with the modeled results to develop a more complete understanding of this electrically heated closed power generation system and to validate the model. The measured and predicted system temperatures and pressures are in good agreement, indicating that the model is a reasonable representation of the test loop. Typical deviations between the model and the hardware results are less than 10%. Additional tests were performed to assess the capability of the Brayton engine to continue to remove decay heat after the reactor/heater is shutdown, to develop safe and effective control strategies, and to access the effectiveness of gas inventory control as an alternative means to provide load following. In one test the heater power was turned off to simulate a rapid reactor shutdown, and the turbomachinery was driven solely by the sensible heat stored in the heater for over 71 minutes without external power input. This is an important safety feature for CBC systems as it means that the closed Brayton loop will keep cooling the reactor without the need for auxiliary power (other than that needed to circulate the waste heat rejection coolant) provided the heat sink is available.

  2. Fault-tolerant reactor protection system

    DOEpatents

    Gaubatz, Donald C.

    1997-01-01

    A reactor protection system having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Each division performs independently of the others (asynchronous operation). All communications between the divisions are asynchronous. Each chassis substitutes its own spare sensor reading in the 2/3 vote if a sensor reading from one of the other chassis is faulty or missing. Therefore the presence of at least two valid sensor readings in excess of a set point is required before terminating the output to the hardware logic of a scram inhibition signal even when one of the four sensors is faulty or when one of the divisions is out of service.

  3. Fault-tolerant reactor protection system

    DOEpatents

    Gaubatz, D.C.

    1997-04-15

    A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Each division performs independently of the others (asynchronous operation). All communications between the divisions are asynchronous. Each chassis substitutes its own spare sensor reading in the 2/3 vote if a sensor reading from one of the other chassis is faulty or missing. Therefore the presence of at least two valid sensor readings in excess of a set point is required before terminating the output to the hardware logic of a scram inhibition signal even when one of the four sensors is faulty or when one of the divisions is out of service. 16 figs.

  4. Development of Pressurized Water Reactor Integrated Safety Analysis Methodology Using Multilevel Coupling Algorithm

    SciTech Connect

    Ziabletsev, Dmitri; Avramova, Maria; Ivanov, Kostadin

    2004-11-15

    The subchannel code COBRA-TF has been introduced for an evaluation of thermal margins on the local pin-by-pin level in a pressurized water reactor. The coupling of COBRA-TF with TRAC-PF1/NEM is performed by providing from TRAC to COBRA-TF axial and radial thermal-hydraulic boundary conditions and relative pin-power profiles, obtained with the pin power reconstruction model of the nodal expansion method (NEM). An efficient algorithm for coupling of the subchannel code COBRA-TF with TRAC-PF1/NEM in the parallel virtual machine environment was developed addressing the issues of time synchronization, data exchange, spatial overlays, and coupled convergence. Local feedback modeling on the pin level was implemented into COBRA-TF, which enabled updating the local form functions and the recalculation of the pin powers in TRAC-PF1/NEM after obtaining the local feedback parameters. The coupled TRAC-PF1/NEM/COBRA-TF code system was tested on the rod ejection accident and main steam line break benchmark problems. In both problems, the local results are closer than before the introduced multilevel coupling to the corresponding critical limits. This fact indicates that the assembly average results tend to underestimate the accident consequences in terms of local safety margins. The capability of local safety evaluation, performed simultaneously (online) with coupled global three-dimensional neutron kinetics/thermal-hydraulic calculations, is introduced and tested. The obtained results demonstrate the importance of the current work.

  5. NASA Aviation Safety Reporting System (ASRS)

    NASA Technical Reports Server (NTRS)

    Connell, Linda

    2011-01-01

    The NASA Aviation Safety Reporting System (ASRS) collects, analyzes, and distributes de-identified safety information provided through confidentially submitted reports from frontline aviation personnel. Since its inception in 1976, the ASRS has collected over 900,000 reports and has never breached the identity of the people sharing their information about events or safety issues. From this volume of data, the ASRS has released over 5,500 aviation safety alerts concerning potential hazards and safety concerns. The ASRS processes these reports, evaluates the information, and provides de-identified report information through the online ASRS Database at http://asrs.arc.nasa.gov. The NASA ASRS is also a founding member of the International Confidential Aviation Safety Systems (ICASS) group which is a collection of other national aviation reporting systems throughout the world. The ASRS model has also been replicated for application to improving safety in railroad, medical, fire fighting, and other domains. This presentation \\vill discuss confidential, voluntary, and non-punitive reporting systems and their advantages in providing information for safety improvements.

  6. Magnetohydrodynamic Vapor Explosions: A Study with Potential Interest to the Safety of Fusion Reactor Project

    NASA Astrophysics Data System (ADS)

    Arias, F. J.

    2010-04-01

    In this paper, the possibility of vapor explosions in superheat liquids in presence of a magnetic field that undergo sudden variation of magnetic field is discussed. This possible phenomenon may play a very important role in the blanket design of future fusion reactors, where transients in magnetic field on liquid metals, could will be a potential hazard for safety.

  7. Two-field and drift-flux models with applications to nuclear reactor safety

    SciTech Connect

    Travis, J.R.

    1984-05-01

    The ideas of the two-field (6 equation model) and drift-flux (4 equation model) description of two-phase flows are presented. Several example calculations relating to reactor safety are discussed and comparisons of the numerical results and experimental data are shown to be in good agreement.

  8. Nuclear plant-aging research on reactor protection systems

    SciTech Connect

    Meyer, L.C.

    1988-01-01

    This report presents the rsults of a review of the Reactor Trip System (RTS) and the Engineered Safety Feature Actuating System (ESFAS) operating experiences reported in Licensee Event Reports (LER)s, the Nuclear Power Experience data base, Nuclear Plant Reliability Data System, and plant maintenance records. Our purpose is to evaluate the potential significance of aging, including cycling, trips, and testing as contributors to degradation of the RTS and ESFAS. Tables are presented that show the percentage of events for RTS and ESFAS classified by cause, components, and subcomponents for each of the Nuclear Steam Supply System vendors. A representative Babcock and Wilcox plant was selected for detailed study. The US Nuclear Regulatory Commission's Nuclear Plant Aging Research guidelines were followed in performing the detailed study that identified materials susceptible to aging, stressors, environmental factors, and failure modes for the RTS and ESFAS as generic instrumentation and control systems. Functional indicators of degradation are listed, testing requirements evaluated, and regulatory issues discussed.

  9. The Safety System of the Herschel Cryostat

    NASA Astrophysics Data System (ADS)

    Langfermann, M.; Jahn, G.; Hohn, R.; Ruehe, W.; Jewell, C.

    2004-06-01

    The cryostat for the `Herschel Space Observatory' for the European Space Agency (ESA) science program, planned for a launch with Ariane 5 in 2007, is designed for 6 days ground hold time and 3.5 years lifetime in orbit. The system comprises two tanks containing about 346 kg of liquid and superfluid Helium, with two cryogenic cold safety valves and burst disks, surrounded by three vapor cooled shields and a vacuum vessel. The safety system is two faults tolerant with three independent paths for pressure relief. The analyses of failure modes and resulting mass flows and the safety elements of the cryogenic system will be discussed.

  10. Analysis of N-16 concentration in primary cooling system of AP1000 power reactor

    NASA Astrophysics Data System (ADS)

    Rohanda, Anis; Waris, Abdul

    2015-04-01

    Nitrogen-16 (N-16) is one of the radiation safety parameter on the primary reactor system. The activation product, N-16, is the predominant contributor to the activity in the reactor coolant system during reactor operation. N-16 is activation product derived from activation of O-16 with fast neutron based on 16O(n,p)16N reaction. Thus study is needed and it performs to determine N-16 concentration in reactor coolant (primary coolant) in supporting radiation safety. One of the way is using analytical methode based on activation and redecay princip to obtain N-16 concentration. The analysis was performed on the configuration basis and operational of Westinghouse AP1000 power reactor in several monitoring points at coolant reactor system. The results of the calculation of N-16 concentration at the core outlet, reactor vessel outlet, pressurizer line, inlet and outlet of steam generators, primary pumps, reactor vessels inlet and core inlet are: 281, 257, 255, 250, 145, 142, 129 and 112 µCi/gram respectively. The results of analysis compared with AP1000 design control document as standard values. The verification showed very high accuracy comparation between analytical results and standard values.

  11. Analysis of N-16 concentration in primary cooling system of AP1000 power reactor

    SciTech Connect

    Rohanda, Anis; Waris, Abdul

    2015-04-16

    Nitrogen-16 (N-16) is one of the radiation safety parameter on the primary reactor system. The activation product, N-16, is the predominant contributor to the activity in the reactor coolant system during reactor operation. N-16 is activation product derived from activation of O-16 with fast neutron based on {sup 16}O(n,p){sup 16}N reaction. Thus study is needed and it performs to determine N-16 concentration in reactor coolant (primary coolant) in supporting radiation safety. One of the way is using analytical methode based on activation and redecay princip to obtain N-16 concentration. The analysis was performed on the configuration basis and operational of Westinghouse AP1000 power reactor in several monitoring points at coolant reactor system. The results of the calculation of N-16 concentration at the core outlet, reactor vessel outlet, pressurizer line, inlet and outlet of steam generators, primary pumps, reactor vessels inlet and core inlet are: 281, 257, 255, 250, 145, 142, 129 and 112 µCi/gram respectively. The results of analysis compared with AP1000 design control document as standard values. The verification showed very high accuracy comparation between analytical results and standard values.

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

    SciTech Connect

    Farmer, Mitchell T.; Bunt, R.; Corradini, M.; Ellison, Paul B.; Francis, M.; Gabor, John D.; Gauntt, R.; Henry, C.; Linthicum, R.; Luangdilok, W.; Lutz, R.; Paik, C.; Plys, M.; Rabiti, Cristian; Rempe, J.; Robb, K.; Wachowiak, R.

    2015-01-31

    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 affect reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).

  13. Integrating system safety into the basic systems engineering process

    NASA Technical Reports Server (NTRS)

    Griswold, J. W.

    1971-01-01

    The basic elements of a systems engineering process are given along with a detailed description of what the safety system requires from the systems engineering process. Also discussed is the safety that the system provides to other subfunctions of systems engineering.

  14. SAFETY SYSTEM FOR CONTROL ROD

    DOEpatents

    Paget, J.A.

    1963-05-14

    A structure for monitoring the structural continuity of a control rod foi a neutron reactor is presented. A electric conductor readily breakable under mechanical stress is fastened along the length of the control rod at a plurality of positions and forms a closed circuit with remote electrical components responsive to an open circuit. A portion of the conductor between the control rod and said components is helically wound to allow free and normally unrestricted movement of the segment of conductor secured to the control rod relative to the remote components. Any break in the circuit is indicative of control rod breakage. (AEC)

  15. Aviation Safety Reporting System: Process and Procedures

    NASA Technical Reports Server (NTRS)

    Connell, Linda J.

    1997-01-01

    The Aviation Safety Reporting System (ASRS) was established in 1976 under an agreement between the Federal Aviation Administration (FAA) and the National Aeronautics and Space Administration (NASA). This cooperative safety program invites pilots, air traffic controllers, flight attendants, maintenance personnel, and others to voluntarily report to NASA any aviation incident or safety hazard. The FAA provides most of the program funding. NASA administers the program, sets its policies in consultation with the FAA and aviation community, and receives the reports submitted to the program. The FAA offers those who use the ASRS program two important reporting guarantees: confidentiality and limited immunity. Reports sent to ASRS are held in strict confidence. More than 350,000 reports have been submitted since the program's beginning without a single reporter's identity being revealed. ASRS removes all personal names and other potentially identifying information before entering reports into its database. This system is a very successful, proof-of-concept for gathering safety data in order to provide timely information about safety issues. The ASRS information is crucial to aviation safety efforts both nationally and internationally. It can be utilized as the first step in safety by providing the direction and content to informed policies, procedures, and research, especially human factors. The ASRS process and procedures will be presented as one model of safety reporting feedback systems.

  16. The secure, transportable, autonomous reactor (STAR): a small proliferation-resistant reactor system for developing countries

    SciTech Connect

    Brown, N W; Hassberger, J A; Smith, C F

    1999-05-27

    The Secure, Transportable, Autonomous Reactor (STAR), is an integrated concept for a small, proliferation-resistant nuclear power system capable of meeting the growing power demands of many regions of the developing world. The STAR approach builds on earlier work investigating the features required for implementation of such a system. The STAR approach includes establishing overall system requirements, conducting research into issues common to four reactor concepts (gas, liquid metal, light water and molten salt), and defining and performing the down-selection to a preferred concept that will serve as the basis for continued development leading to an eventual prototype. The paper indicates that a number of unique and distinguishing innovations are needed to both meet the energy demands of most of the world's developing regions and address growing nuclear proliferation concerns. These technical innovations form much of the basis underlying the STAR concept and include: eliminating on-site refueling and fuel access; incorporating a systems approach to nuclear energy supply and infrastructure design, with all aspects of equipment life, fuel and waste cycles included; small unit size enabling transportability; replaceable standardized modular design; resilient and robust design concepts leading to large safety margins, high reliability and reduced maintenance; simplicity in operation with reliance on autonomous control and remote monitoring; and waste minimization and waste form optimization.

  17. NASA Aviation Safety Reporting System

    NASA Technical Reports Server (NTRS)

    1980-01-01

    Problems in briefing of relief by air traffic controllers are discussed, including problems that arise when duty positions are changed by controllers. Altimeter reading and setting errors as factors in aviation safety are discussed, including problems associated with altitude-including instruments. A sample of reports from pilots and controllers is included, covering the topics of ATIS broadcasts an clearance readback problems. A selection of Alert Bulletins, with their responses, is included.

  18. Containment system for supercritical water oxidation reactor

    SciTech Connect

    Chastagner, P.

    1991-12-31

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

  19. Containment system for supercritical water oxidation reactor

    DOEpatents

    Chastagner, Philippe

    1994-01-01

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

  20. Containment system for supercritical water oxidation reactor

    DOEpatents

    Chastagner, P.

    1994-07-05

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

  1. Common-Cause Failure Analysis for Reactor Protection System Reliability Studies

    SciTech Connect

    Gentillon, C.; Rasmuson, D.; Eide, S.; Wierman, T.

    1999-08-01

    Analyses were performed of the safety-related performance of the reactor protection system (RPS) at U.S. Westinghouse and General Electric commercial reactors during the period 1984 through 1995. RPS operational data from these reactors were collected from the Nuclear Plant Reliability Data System (NPRDS) and Licensee Event Reports (LER). The common-cause failure (CCF) modeling in the fault trees developed for these studies and the analysis and use of common-cause failure data were sophisticated, state-of-the-art efforts. The overall CCF effort helped to test and expand the limits of the U.S. Nuclear Regulatory Commission's CCF methodology.

  2. RISMC advanced safety analysis working plan: FY2015 - FY2019. Light Water Reactor Sustainability Program

    SciTech Connect

    Szilard, Ronaldo H; Smith, Curtis L

    2014-09-01

    In this report, the Advanced Safety Analysis Program (ASAP) objectives and value proposition is described. ASAP focuses on modernization of nuclear power safety analysis (tools, methods and data); implementing state-of-the-art modeling techniques (which include, for example, enabling incorporation of more detailed physics as they become available); taking advantage of modern computing hardware; and combining probabilistic and mechanistic analyses to enable a risk informed safety analysis process. The modernized tools will maintain the current high level of safety in our nuclear power plant fleet, while providing an improved understanding of safety margins and the critical parameters that affect them. Thus, the set of tools will provide information to inform decisions on plant modifications, refurbishments, and surveillance programs, while improving economics. The set of tools will also benefit the design of new reactors, enhancing safety per unit cost of a nuclear plant. As part of the discussion, we have identified three sets of stakeholders, the nuclear industry, the Department of Energy (DOE), and associated oversight organizations. These three groups would benefit from ASAP in different ways. For example, within the DOE complex, the possible applications that are seen include the safety of experimental reactors, facility life extension, safety-by-design in future generation advanced reactors, and managing security for the storage of nuclear material. This report provides information in five areas: (1) A value proposition (“why is this important?”) that will make the case for stakeholder’s use of the ASAP research and development (R&D) products; (2) An identification of likely end users and pathway to adoption of enhanced tools by the end-users; (3) A proposed set of practical and achievable “use case” demonstrations; (4) A proposed plan to address ASAP verification and validation (V&V) needs; and (5) A proposed schedule for the multi-year ASAP.

  3. Fast reactor safety: proceedings of the international topical meeting. Volume 2. [R

    SciTech Connect

    Not Available

    1985-07-01

    The emphasis of this meeting was on the safety-related aspects of fast reactor design, analysis, licensing, construction, and operation. Relative to past meetings, there was less emphasis on the scientific and technological basis for accident assessment. Because of its broad scope, the meeting attracted 217 attendees from a wide cross section of the design, safety analysis, and safety technology communities. Eight countries and two international organizations were represented. A total of 126 papers were presented, with contributions from the United States, France, Japan, the United Kingdom, Germany, and Italy. Sessions covered in Volume 2 include: safety design concepts; operational transient experiments; analysis of seismic and external events; HCDA-related codes, analysis, and experiments; sodium fires; instrumentation and control/PPS design; whole-core accident analysis codes; and impact of safety design considerations on future LMFBR developments.

  4. Code development incorporating environmental, safety, and economic aspects of fusion reactors (FY 92--94). Final report

    SciTech Connect

    Ho, S.K.; Fowler, T.K.; Holdren, J.P.

    1994-11-01

    This is the Final Report for a three-year (FY 92--94) study of the Environmental, Safety, and Economic (ESE) aspects of fusion energy systems, emphasizing development of computerized approaches suitable for incorporation as modules in fusion system design codes. First, as is reported in Section 2, the authors now have operating a simplified but complete environment and safety evaluation code, BESAFE. The first tests of BESAFE as a module of the SUPERCODE, a design optimization systems code at LLNL, are reported in Section 3. Secondly, as reported in Section 4, the authors have maintained a strong effort in developing fast calculational schemes for activation inventory evaluation. In addition to these major accomplishments, considerable progress has been made on research on specific topics as follows. A tritium modeling code TRIDYN was developed in collaboration with the TSTA group at LANL and the Fusion Nuclear Technology group at UCLA. A simplified algorithm has been derived to calculate the transient temperature profiles in the blanket during accidents. The scheme solves iteratively a system of non-linear ordinary differential equations describing about 10 regions of the blanket by preserving energy balance. The authors have studied the physics and engineering aspects of divertor modeling for safety applications. Several modifications in the automation and characterization of environmental and safety indices have been made. They have applied this work to the environmental and safety comparisons of stainless steel with alternative structural materials for fusion reactors. A methodology in decision analysis utilizing influence and decision diagrams has been developed to model fusion reactor design problems. Most of the work during this funding period has been reported in 26 publications including theses, journal publications, conference papers, and technical reports, as listed in Section 11.

  5. Static conversion systems. [for space power reactors

    NASA Technical Reports Server (NTRS)

    Ewell, R.; Mondt, J.

    1985-01-01

    Historically, all space power systems that have actually flown in space have relied on static energy conversion technology. Thus, static conversion is being considered for space nuclear power systems as well. There are four potential static conversion technologies which should be considered. These include: the alkali metal thermoelectric converter (AMTEC), the thermionic converter, the thermoelectric converter, and the thermophotovoltaic converter (TPV). These four conversion technologies will be described in brief detail along with their current status and development needs. In addition, the systems implications of using each of these conversion technologies with a space nuclear reactor power system will be evaluated and some comparisons made.

  6. 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

  7. Staged membrane oxidation reactor system

    DOEpatents

    Repasky, John Michael; Carolan, Michael Francis; Stein, VanEric Edward; Chen, Christopher Ming-Poh

    2014-05-20

    Ion transport membrane oxidation system comprising (a) two or more membrane oxidation stages, each stage comprising a reactant zone, an oxidant zone, one or more ion transport membranes separating the reactant zone from the oxidant zone, a reactant gas inlet region, a reactant gas outlet region, an oxidant gas inlet region, and an oxidant gas outlet region; (b) an interstage reactant gas flow path disposed between each pair of membrane oxidation stages and adapted to place the reactant gas outlet region of a first stage of the pair in flow communication with the reactant gas inlet region of a second stage of the pair; and (c) one or more reactant interstage feed gas lines, each line being in flow communication with any interstage reactant gas flow path or with the reactant zone of any membrane oxidation stage receiving interstage reactant gas.

  8. Staged membrane oxidation reactor system

    DOEpatents

    Repasky, John Michael; Carolan, Michael Francis; Stein, VanEric Edward; Chen, Christopher Ming-Poh

    2013-04-16

    Ion transport membrane oxidation system comprising (a) two or more membrane oxidation stages, each stage comprising a reactant zone, an oxidant zone, one or more ion transport membranes separating the reactant zone from the oxidant zone, a reactant gas inlet region, a reactant gas outlet region, an oxidant gas inlet region, and an oxidant gas outlet region; (b) an interstage reactant gas flow path disposed between each pair of membrane oxidation stages and adapted to place the reactant gas outlet region of a first stage of the pair in flow communication with the reactant gas inlet region of a second stage of the pair; and (c) one or more reactant interstage feed gas lines, each line being in flow communication with any interstage reactant gas flow path or with the reactant zone of any membrane oxidation stage receiving interstage reactant gas.

  9. Staged membrane oxidation reactor system

    DOEpatents

    Repasky, John Michael; Carolan, Michael Francis; Stein, VanEric Edward; Chen, Christopher Ming-Poh

    2012-09-11

    Ion transport membrane oxidation system comprising (a) two or more membrane oxidation stages, each stage comprising a reactant zone, an oxidant zone, one or more ion transport membranes separating the reactant zone from the oxidant zone, a reactant gas inlet region, a reactant gas outlet region, an oxidant gas inlet region, and an oxidant gas outlet region; (b) an interstage reactant gas flow path disposed between each pair of membrane oxidation stages and adapted to place the reactant gas outlet region of a first stage of the pair in flow communication with the reactant gas inlet region of a second stage of the pair; and (c) one or more reactant interstage feed gas lines, each line being in flow communication with any interstage reactant gas flow path or with the reactant zone of any membrane oxidation stage receiving interstage reactant gas.

  10. Operating experience feedback report: Reliability of safety-related steam turbine-driven standby pumps. Commercial power reactors, Volume 10

    SciTech Connect

    Boardman, J.R.

    1994-10-01

    This report documents a detailed analysis of failure initiators, causes and design features for steam turbine assemblies (turbines with their related components, such as governors and valves) which are used as drivers for standby pumps in the auxiliary feedwater systems of US commercial pressurized water reactor plants, and in the high pressure coolant injection and reactor core isolation cooling systems of US commercial boiling water reactor plants. These standby pumps provide a redundant source of water to remove reactor core heat as specified in individual plant safety analysis reports. The period of review for this report was from January 1974 through December 1990 for licensee event reports (LERS) and January 1985 through December 1990 for Nuclear Plant Reliability Data System (NPRDS) failure data. This study confirmed the continuing validity of conclusions of earlier studies by the US Nuclear Regulatory Commission and by the US nuclear industry that the most significant factors in failures of turbine-driven standby pumps have been the failures of the turbine-drivers and their controls. Inadequate maintenance and the use of inappropriate vendor technical information were identified as significant factors which caused recurring failures.

  11. Safety assessment of the tritium recovery system

    SciTech Connect

    Not Available

    1994-09-01

    This Safety Assessment (SA) contains descriptions and evaluations of the environmental, health, and safety issues associated with the Tritium Recovery System (TRS) at the Pinellas Plant. It provides: 1. site and facility descriptions, 2. an overall description of the TRS and its operations, 3. an evaluation of the hazards associated with TRS operations, 4. descriptions and analyses of the adequacy of measures taken to eliminate, control, or mitigate identified hazards, and 5. assessment of potential accidents and their associated risks. This SA contains the results of safety evaluations of TRS operations, equipment, and supplied systems. The evaluations include, as appropriate, preliminary hazards listings, qualitative risk assessments, and quantitative risk assessments.

  12. Autonomous Flight Safety System - Phase III

    NASA Technical Reports Server (NTRS)

    2008-01-01

    The Autonomous Flight Safety System (AFSS) is a joint KSC and Wallops Flight Facility project that uses tracking and attitude data from onboard Global Positioning System (GPS) and inertial measurement unit (IMU) sensors and configurable rule-based algorithms to make flight termination decisions. AFSS objectives are to increase launch capabilities by permitting launches from locations without range safety infrastructure, reduce costs by eliminating some downrange tracking and communication assets, and reduce the reaction time for flight termination decisions.

  13. CLASSIFICATION OF THE MGR HEALTH SAFETY SYSTEM

    SciTech Connect

    J.A. Ziegler

    1999-08-31

    The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) health safety system structures, systems and components (SSCs) performed by the MGR Safety Assurance Department. This analysis also provides the basis for revision of YMP/90-55Q, Q-List (YMP 1998). The Q-List identifies those MGR SSCs subject to the requirements of DOE/RW-0333P, ''Quality Assurance Requirements and Description'' (QARD) (DOE 1998).

  14. Integral reactor system and method for fuel cells

    SciTech Connect

    Fernandes, Neil Edward; Brown, Michael S; Cheekatamarla, Praveen; Deng, Thomas; Dimitrakopoulos, James; Litka, Anthony F

    2013-11-19

    A reactor system is integrated internally within an anode-side cavity of a fuel cell. The reactor system is configured to convert hydrocarbons to smaller species while mitigating the lower production of solid carbon. The reactor system may incorporate one or more of a pre-reforming section, an anode exhaust gas recirculation device, and a reforming section.

  15. Microchannel Reactor System for Catalytic Hydrogenation

    SciTech Connect

    Adeniyi Lawal; Woo Lee; Ron Besser; Donald Kientzler; Luke Achenie

    2010-12-22

    We successfully demonstrated a novel process intensification concept enabled by the development of microchannel reactors, for energy efficient catalytic hydrogenation reactions at moderate temperature, and pressure, and low solvent levels. We designed, fabricated, evaluated, and optimized a laboratory-scale microchannel reactor system for hydrogenation of onitroanisole and a proprietary BMS molecule. In the second phase of the program, as a prelude to full-scale commercialization, we designed and developed a fully-automated skid-mounted multichannel microreactor pilot plant system for multiphase reactions. The system is capable of processing 1 – 10 kg/h of liquid substrate, and an industrially relevant immiscible liquid-liquid was successfully demonstrated on the system. Our microreactor-based pilot plant is one-of-akind. We anticipate that this process intensification concept, if successfully demonstrated, will provide a paradigm-changing basis for replacing existing energy inefficient, cost ineffective, environmentally detrimental slurry semi-batch reactor-based manufacturing practiced in the pharmaceutical and fine chemicals industries.

  16. 49 CFR 659.19 - System safety program plan: contents.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 49 Transportation 7 2014-10-01 2014-10-01 false System safety program plan: contents. 659.19... State Oversight Agency § 659.19 System safety program plan: contents. The system safety plan shall... safety program and describes the authority that establishes the system safety program plan. (b) A...

  17. 49 CFR 659.19 - System safety program plan: contents.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 49 Transportation 7 2012-10-01 2012-10-01 false System safety program plan: contents. 659.19... State Oversight Agency § 659.19 System safety program plan: contents. The system safety plan shall... safety program and describes the authority that establishes the system safety program plan. (b) A...

  18. Plasma generators, reactor systems and related methods

    DOEpatents

    Kong, Peter C.; Pink, Robert J.; Lee, James E.

    2007-06-19

    A plasma generator, reactor and associated systems and methods are provided in accordance with the present invention. A plasma reactor may include multiple sections or modules which are removably coupled together to form a chamber. Associated with each section is an electrode set including three electrodes with each electrode being coupled to a single phase of a three-phase alternating current (AC) power supply. The electrodes are disposed about a longitudinal centerline of the chamber and are arranged to provide and extended arc and generate an extended body of plasma. The electrodes are displaceable relative to the longitudinal centerline of the chamber. A control system may be utilized so as to automatically displace the electrodes and define an electrode gap responsive to measure voltage or current levels of the associated power supply.

  19. Reactor safety issues resolved by the 2D/3D Program. International Agreement Report

    SciTech Connect

    Damerell, P.S.; Simons, J.W.

    1993-07-01

    The 2D/3D Program studied multidimensional thermal-hydraulics in a PWR core and primary system during the end-of-blowdown and post-blowdown phases of a large-break LOCA (LBLOCA), and during selected small-break LOCA (SBLOCA) transients. The program included tests at the Cylindrical Core Test Facility (CCTF), the Slab Core Test Facility (SCTF), and the Upper Plenum Test Facility (UPTF), and computer analyses using TRAC. Tests at CCTF investigated core thermal-hydraulics and overall system behavior while tests at SCTF concentrated on multidimensional core thermal-hydraulics. The UPTF tests investigated two-phase flow behavior in the downcomer, upper plenum, tie plate region, and primary loops. TRAC analyses evaluated thermal-hydraulic behavior throughout the primary system in tests as well as in PWRs. This report summarizes the test and analysis results in each of the main areas where improved information was obtained in the 2D/3D Program. The discussion is organized in terms of the reactor safety issues investigated.

  20. Application of system safety to rail transit systems

    NASA Technical Reports Server (NTRS)

    Styles, T. D.

    1971-01-01

    Management emphasis on system safety in the rapid transit industry includes the granting and use of funds by the Federal Government according to systematic analysis of safety hazards in advance. Likelihood predictions that those hazards will be activated by exposure of the system to a system failure, a human error, external conditions, or combinations of these aspects determine alternatives to the assumption of risk and recommend corrections before the system is operational. Rigorous safety analyses are projected to assure operational safety for prolonged periods under varied maintenance conditions; these analysis encompass station accident possibilities as well as train-person collisions, car equipment and design, traffic control systems, and tunnel design problems.

  1. Power conditioning for space nuclear reactor systems

    NASA Technical Reports Server (NTRS)

    Berman, Baruch

    1987-01-01

    This paper addresses the power conditioning subsystem for both Stirling and Brayton conversion of space nuclear reactor systems. Included are the requirements summary, trade results related to subsystem implementation, subsystem description, voltage level versus weight, efficiency and operational integrity, components selection, and shielding considerations. The discussion is supported by pertinent circuit and block diagrams. Summary conclusions and recommendations derived from the above studies are included.

  2. Systems pharmacology augments drug safety surveillance.

    PubMed

    Lorberbaum, T; Nasir, M; Keiser, M J; Vilar, S; Hripcsak, G; Tatonetti, N P

    2015-02-01

    Small molecule drugs are the foundation of modern medical practice, yet their use is limited by the onset of unexpected and severe adverse events (AEs). Regulatory agencies rely on postmarketing surveillance to monitor safety once drugs are approved for clinical use. Despite advances in pharmacovigilance methods that address issues of confounding bias, clinical data of AEs are inherently noisy. Systems pharmacology-the integration of systems biology and chemical genomics-can illuminate drug mechanisms of action. We hypothesize that these data can improve drug safety surveillance by highlighting drugs with a mechanistic connection to the target phenotype (enriching true positives) and filtering those that do not (depleting false positives). We present an algorithm, the modular assembly of drug safety subnetworks (MADSS), to combine systems pharmacology and pharmacovigilance data and significantly improve drug safety monitoring for four clinically relevant adverse drug reactions. PMID:25670520

  3. The Global Maritime Distress and Safety System

    NASA Astrophysics Data System (ADS)

    Kent, Peter E.

    1990-08-01

    The Global Maritime Distress and Safety System (GMDSS) is discussed with respect to its initial planning, the communication network, and other details, including the fully automated Maritime Safety Information service and the implementation of the whole system. GMDSS is the result of international cooperation over a period of about 10 years and provides the maritime community with an integrated distress and safety communication system which significantly enhances the safety of life and property in the harsh environment of the sea. Probably the most essential element of the GMDSS is the provision of an adequate communication network which will permit ships in need of assistance to notify responsible authorities, discuss the help they need, and allow the search and rescue activities to be coordinated by the most appropriate center.

  4. Advanced Test Reactor (ATR) Facility 10CFR830 Safety Basis Related to Facility Experiments

    SciTech Connect

    Tomberlin, Terry Alan

    2002-06-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) Advanced Test Reactor (ATR), a DOE Category A reactor, was designed to provide an irradiation test environment for conducting a variety of experiments. The ATR Safety Analysis Report, determined by DOE to meet the requirements of 10 CFR 830, Subpart B, provides versatility in types of experiments that may be conducted. This paper addresses two general types of experiments in the ATR facility and how safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore this type of experiment is addressed with more detail in the safety basis. This allows individual safety analyses for these experiments to be more routine and repetitive. The second type of experiment is less defined and is permitted under more general controls. Therefore, individual safety analyses for the second type of experiment tend to be more unique from experiment to experiment. Experiments are also discussed relative to "major modifications" and DOE-STD-1027-92. Application of the USQ process to ATR experiments is also discussed.

  5. Advanced Test Reactor (ATR) Facility 10CFR830 Safety Basis Related to Facility Experiments

    SciTech Connect

    Tomberlin, T.A.

    2002-06-19

    The Idaho National Engineering and Environmental Laboratory (INEEL) Advanced Test Reactor (ATR), a DOE Category A reactor, was designed to provide an irradiation test environment for conducting a variety of experiments. The ATR Safety Analysis Report, determined by DOE to meet the requirements of 10 CFR 830, Subpart B, provides versatility in types of experiments that may be conducted. This paper addresses two general types of experiments in the ATR facility and how safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore this type of experiment is addressed with more detail in the safety basis. This allows individual safety analyses for these experiments to be more routine and repetitive. The second type of experiment is less defined and is permitted under more general controls. Therefore, individual safety analyses for the second type of experiment tend to be more unique from experiment to experiment. Experiments are also discussed relative to ''major modifications'' and DOE-STD-1027-92. Application of the USQ process to ATR experiments is also discussed.

  6. Generic safety insights for inspection of boiling water reactors

    SciTech Connect

    Higgins, J.C.; Taylor, J.H.; Fresco, A.N.; Hillman, B.M.

    1987-01-01

    As the number of operating nuclear power plants (NPP) increases, safety inspection has increased in importance. However, precisely what is important, and what is not important. What should one focus inspection efforts on. Over the last two years Probabilistic Risk Assessment (PR) techniques have been developed to aid in the inspection process. Broad interest in generic PRA-based methods has arisen in the past year, since only about 25% of the US nuclear power plants have completed PRAs, and also, inspectors want PRA-based tools for these plants. This paper describes the BNL program to develop generic BWR PRA-based inspection insights or inspection guidance designed to be applied to plants without PRAs.

  7. Reactor component inventory system at FFTF

    SciTech Connect

    Ordonez, C.R.; Redekopp, R.D.; Reed, E.A.

    1985-02-01

    A reliable inventory control system was developed at the Fast Flux Test Facility (FFTF) to keep track of the occupancy of 900 refueling facility locations, to compile historical data on the movement of each reactor assembly, and to simulate assembly moves. The simulate capability is valuable because it allows verification of documents before they are issued for use in the plant, and eliminates the possibility of planning illegal or impossible moves. The system is installed on a UNIVAC 1100 computer and is maintained using a data base management system by Sperry Univac called MAPPER.

  8. System for controlling child safety seat environment

    NASA Technical Reports Server (NTRS)

    Dabney, Richard W. (Inventor); Elrod, Susan V. (Inventor)

    2008-01-01

    A system is provided to control the environment experienced by a child in a child safety seat. Each of a plurality of thermoelectric elements is individually controllable to be one of heated and cooled relative to an ambient temperature. A first portion of the thermoelectric elements are positioned on the child safety seat such that a child sitting therein is positioned thereover. A ventilator coupled to the child safety seat moves air past a second portion of the thermoelectric elements and filters the air moved therepast. One or more jets coupled to the ventilator receive the filtered air. Each jet is coupled to the child safety seat and can be positioned to direct the heated/cooled filtered air to the vicinity of the head of the child sitting in the child safety seat.

  9. Gas-cooled fast breeder reactor. Quarterly progress report, February 1-April 30, 1980

    SciTech Connect

    Not Available

    1980-05-01

    Information is presented concerning the reactor vessel; reactivity control mechanisms and instrumentation; reactor internals; primary coolant circuits;core auxiliary cooling system; reactor core; systems engineering; and reactor safety and reliability;

  10. Reactor physics and safety aspects of various design options of a Russian light water reactor with rock-like fuels

    NASA Astrophysics Data System (ADS)

    Bondarenko, A. V.; Komissarov, O. V.; Kozmenkov, Ya. K.; Matveev, Yu. V.; Orekhov, Yu. I.; Pivovarov, V. A.; Sharapov, V. N.

    2003-06-01

    This paper presents results of analytical studies on weapons grade plutonium incineration in VVER (640) medium size light water reactors using a special composition of rock-like fuel (ROX-fuel) to assure spent fuel long-term storage without its reprocessing. The main goal is to achieve high degree of plutonium incineration in once-through cycle. In this paper we considered two fuel compositions. In both compositions weapons grade plutonium is used as fissile material. Spinel (MgAl 2O 4) is used as the 'preserving' material assuring safe storage of the spent fuel. Besides an inert matrix, the option of rock-like fuel with thorium dioxide was studied. One of principal problems in the realization of the proposed approach is the substantial change of properties of the light water reactor core when passing to the use of the ROX-fuel, in particular: (i) due to the absence of 238U the Doppler effect playing a crucial role in reactor's self-regulation and limiting the consequences of reactivity accidents, decreases significantly, (ii) no fuel breeding on one hand, and the quest to attain the maximum plutonium burnup on the other hand, would result in a drastical change of the fuel assembly power during the lifetime and, as a consequence, the rise in irregularity of the power density of fuel assemblies, (iii) both the control rods worth and dissolved boron worth decrease in view of neutron spectrum hardening brought on by the larger absorption cross-section of plutonium as compared to uranium, (iv) βeff is markedly reduced. All these distinctive features are potentially detrimental to the reactor nuclear safety. The principal objective of this work is that to identify a variant of the fuel composition and the reactor layout, which would permit neutralize the negative effect of the above-mentioned distinctive features.

  11. Supplementary safety system 1/4 scale testing

    SciTech Connect

    Garrett, R.L.; Paik, I.K.

    1993-09-01

    During the course of updating the K-Reactor Safety Analysis Report Chapter 15 in 1990, it was identified that the current Supplementary Safety System (SSS) may not be adequate in protecting the reactor during the process water pump coastdown initiated by a loss of AC power when the safety rods are assumed to fail. A SSS modification project was initiated to add an additional ink injection pathway near the pump suction. In addition, the Department of Energy raised a question on the thermal buoyancy effects on moderator flow pattern and ink dispersion in the moderator space. The development and documentation of a two-dimensional code called MODFLOW was undertaken to describe the problem. This report discusses the results of the moderator flow and ink (Gadolinium Poison Solution - GPS) dispersion tests designed to provide qualified data for validation and benchmarking of the MODFLOW computer code with the secondary objectives being the development of concentration profiles and video footage of simulated GPS dispersion under steady-state and transient flow conditions.

  12. Dynamics and inherent safety features of small modular high temperature gas-cooled reactors

    SciTech Connect

    Harrington, R.M.; Ball, S.J.; Cleveland, J.C.

    1986-01-01

    Investigations were made at Oak Ridge National Laboratory to characterize the dynamics and inherent safety features of various modular high temperature gas-cooled reactor (HTGR) designs. This work was sponsored by the US Nuclear Regulatory Commission's HTGR Safety Research program. The US Department of Energy (DOE) and the Gas Cooled Reactor Associates (GCRA) have sponsored studies of several modular HTGR concepts, each having it own unique advantageous economic and inherent safety features. The DOE design team has recently choses a 350-MW(t) annular core with prismatic, graphite matrix fuel for its reference plant. The various safety features of this plant and of the pebble-bed core designs similar to those currently being developed and operated in the Federal Republic of Germany (FRG) are described. A varity of postulated accident sequences involving combinations of loss of forced circulation of the helium primary coolant, loss of primary coolant pressurization, and loss of normal and backup heat sinks were studied and are discussed. Results demonstrate that each concept can withstand an uncontrolled heatup accident without reaching excessive peak fuel temperatures. Comparisons of calculated and measured response for a loss of forced circulation test on the FRG reactor, AVR, are also presented. 10 refs.

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

    SciTech Connect

    Memmott, M. J.; Stansbury, C.; Taylor, C.

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

  14. Window-closing safety system

    DOEpatents

    McEwan, T.E.

    1997-08-26

    A safety device includes a wire loop embedded in the glass of a passenger car window and routed near the closing leading-edge of the window. The wire loop carries microwave pulses around the loop to and from a transceiver with separate output and input ports. An evanescent field only an inch or two in radius is created along the wire loop by the pulses. Just about any object coming within the evanescent field will dramatically reduce the energy of the microwave pulses received back by the transceiver. Such a loss in energy is interpreted as a closing area blockage, and electrical interlocks are provided to halt or reverse a power window motor that is actively trying to close the window. 5 figs.

  15. Window-closing safety system

    DOEpatents

    McEwan, Thomas E.

    1997-01-01

    A safety device includes a wire loop embedded in the glass of a passenger car window and routed near the closing leading-edge of the window. The wire loop carries microwave pulses around the loop to and from a transceiver with separate output and input ports. An evanescent field only and inch or two in radius is created along the wire loop by the pulses. Just about any object coming within the evanescent field will dramatically reduce the energy of the microwave pulses received back by the transceiver. Such a loss in energy is interpreted as a closing area blockage, and electrical interlocks are provided to halt or reverse a power window motor that is actively trying to close the window.

  16. K West integrated water treatment system subproject safety analysis document

    SciTech Connect

    SEMMENS, L.S.

    1999-02-24

    This Accident Analysis evaluates unmitigated accident scenarios, and identifies Safety Significant and Safety Class structures, systems, and components for the K West Integrated Water Treatment System.

  17. NASA System Safety Handbook. Volume 2: System Safety Concepts, Guidelines, and Implementation Examples

    NASA Technical Reports Server (NTRS)

    Dezfuli, Homayoon; Benjamin, Allan; Everett, Christopher; Feather, Martin; Rutledge, Peter; Sen, Dev; Youngblood, Robert

    2015-01-01

    This is the second of two volumes that collectively comprise the NASA System Safety Handbook. Volume 1 (NASASP-210-580) was prepared for the purpose of presenting the overall framework for System Safety and for providing the general concepts needed to implement the framework. Volume 2 provides guidance for implementing these concepts as an integral part of systems engineering and risk management. This guidance addresses the following functional areas: 1.The development of objectives that collectively define adequate safety for a system, and the safety requirements derived from these objectives that are levied on the system. 2.The conduct of system safety activities, performed to meet the safety requirements, with specific emphasis on the conduct of integrated safety analysis (ISA) as a fundamental means by which systems engineering and risk management decisions are risk-informed. 3.The development of a risk-informed safety case (RISC) at major milestone reviews to argue that the systems safety objectives are satisfied (and therefore that the system is adequately safe). 4.The evaluation of the RISC (including supporting evidence) using a defined set of evaluation criteria, to assess the veracity of the claims made therein in order to support risk acceptance decisions.

  18. Critical review of the reactor-safety study radiological health effects model. Final report

    SciTech Connect

    Cooper, D.W.; Evans, J.S.; Jacob, N.; Kase, K.R.; Maletskos, C.J.; Robertson, J.B.; Smith, D.G.

    1983-03-01

    This review of the radiological health effects models originally presented in the Reactor Safety Study (RSS) and currently used by the US Nuclear Regulatory Commission (NRC) was undertaken to assist the NRC in determining whether or not to revise the models and to aid in the revision, if undertaken. The models as presented in the RSS and as implemented in the CRAC (Calculations of Reactor Accident Consequences) Code are described and critiqued. The major elements analyzed are those concerning dosimetry, early effects, and late effects. The published comments on the models are summarized, as are the important findings since the publication of the RSS.

  19. HRIBF Tandem Accelerator Radiation Safety System Upgrade

    SciTech Connect

    Blankenship, J.L.; Juras, R.C.

    1998-11-04

    The HRIBF Tandem Accelerator Radiation Safety System was designed to permit experimenters and operations staff controlled access to beam transport and experiment areas with accelerated beam present. Neutron-Gamma detectors are mounted in eaeh area at points of maximum dose rate and the resulting signals are integrated by redundan~ circuitry; beam is stopped if dose rate or integrated dose exceeds established limits. This paper will describe the system, in use for several vears at the HRIBF, and discuss changes recently made to modernize the system and to make the system compliant with DOE Order 5480.25 and related ORNL updated safety rules.

  20. HRIBF Tandem Accelerator Radiation Safety System Upgrade

    NASA Astrophysics Data System (ADS)

    Juras, R. C.; Blankenship, J. L.

    1999-06-01

    The HRIBF Tandem Accelerator Radiation Safety System was designed to permit experimenters and operations staff controlled access to beam transport and experiment areas with accelerated beam present. Neutron-Gamma detectors are mounted in each area at points of maximum dose rate and the resulting signals are integrated by redundant circuitry; beam is stopped if dose rate or integrated dose exceeds established limits. This paper will describe the system, in use for several years at the HRIBF, and discuss changes recently made to modernize the system and to make the system compliant with DOE Order 5480.25 and related ORNL updated safety rules.

  1. System Safety in an IT Service Organization

    NASA Astrophysics Data System (ADS)

    Parsons, Mike; Scutt, Simon

    Within Logica UK, over 30 IT service projects are considered safetyrelated. These include operational IT services for airports, railway infrastructure asset management, nationwide radiation monitoring and hospital medical records services. A recent internal audit examined the processes and documents used to manage system safety on these services and made a series of recommendations for improvement. This paper looks at the changes and the challenges to introducing them, especially where the service is provided by multiple units supporting both safety and non-safety related services from multiple locations around the world. The recommendations include improvements to service agreements, improved process definitions, routine safety assessment of changes, enhanced call logging, improved staff competency and training, and increased safety awareness. Progress is reported as of today, together with a road map for implementation of the improvements to the service safety management system. A proposal for service assurance levels (SALs) is discussed as a way forward to cover the wide variety of services and associated safety risks.

  2. Documented Safety Analysis Addendum for the Neutron Radiography Reactor Facility Core Conversion

    SciTech Connect

    Boyd D. Christensen

    2009-05-01

    The Neutron Radiography Reactor Facility (NRAD) is a Training, Research, Isotope Production, General Atomics (TRIGA) reactor which was installed in the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) in the mid 1970s. The facility provides researchers the capability to examine both irradiated and non-irradiated materials in support of reactor fuel and components programs through non-destructive neutron radiography examination. The facility has been used in the past as one facet of a suite of reactor fuels and component examination facilities available to researchers at the INL and throughout the DOE complex. The facility has also served various commercial research activities in addition to the DOE research and development support. The reactor was initially constructed using Fuel Lifetime Improvement Program (FLIP)- type highly enriched uranium (HEU) fuel obtained from the dismantled Puerto Rico Nuclear Center (PRNC) reactor. In accordance with international non-proliferation agreements, the NRAD core will be converted to a low enriched uranium (LEU) fuel and will continue to utilize the PRNC control rods, control rod drives, startup source, and instrument console as was previously used with the HEU core. The existing NRAD Safety Analysis Report (SAR) was created and maintained in the preferred format of the day, combining sections of both DOE-STD-3009 and Nuclear Regulatory Commission Regulatory Guide 1.70. An addendum was developed to cover the refueling and reactor operation with the LEU core. This addendum follows the existing SAR format combining required formats from both the DOE and NRC. This paper discusses the project to successfully write a compliant and approved addendum to the existing safety basis documents.

  3. Nuclear reactor pressure vessel support system

    DOEpatents

    Sepelak, George R.

    1978-01-01

    A support system for nuclear reactor pressure vessels which can withstand all possible combinations of stresses caused by a postulated core disrupting accident during reactor operation. The nuclear reactor pressure vessel is provided with a flange around the upper periphery thereof, and the flange includes an annular vertical extension formed integral therewith. A support ring is positioned atop of the support ledge and the flange vertical extension, and is bolted to both members. The plug riser is secured to the flange vertical extension and to the top of a radially outwardly extension of the rotatable plug. This system eliminates one joint through which fluids contained in the vessel could escape by making the fluid flow path through the joint between the flange and the support ring follow the same path through which fluid could escape through the plug risers. In this manner, the sealing means to prohibit the escape of contained fluids through the plug risers can also prohibit the escape of contained fluid through the securing joint.

  4. Consumer product safety: A systems problem

    NASA Technical Reports Server (NTRS)

    Clark, C. C.

    1971-01-01

    The manufacturer, tester, retailer, consumer, repairer disposer, trade and professional associations, national and international standards bodies, and governments in several roles are all involved in consumer product safety. A preliminary analysis, drawing on system safety techniques, is utilized to distinguish the inter-relations of these many groups and the responsibilities that they are or could take for product safety, including the slow accident hazards as well as the more commonly discussed fast accident hazards. The importance of interactive computer aided information flow among these groups is particularly stressed.

  5. (Instrumentation and controls technology and reactor operational safety)

    SciTech Connect

    White, J.D.

    1990-12-17

    While on vacation, the traveler participated as a co-chairman of a panel of instrumentation and controls specialists visiting nuclear establishments in Europe. The purpose of the visit was to assess the status of instrumentation and controls technology for nuclear power in Europe. A list of the sites visited and the personnel contacted is included in this trip report. The visit was sponsored by Loyola College working under contract to the National Science Foundation. All costs were paid by Loyola College, for whom the traveler was a consultant. This was an outside activity approved by DOE. The traveler was surprised by the high level of automaton present in the German Konvoi nuclear power plants built by Siemens AG KWU. The claim was that this was done to improve the safety of the plant by keeping the operator out of the loop'' for the first 30 minutes of some transients or accidents. The traveler was also surprised by the high level of man-machine interface R D in the USSR.

  6. Nuclear reactor insulation and preheat system

    DOEpatents

    Wampole, Nevin C.

    1978-01-01

    An insulation and preheat system for preselected components of a fluid cooled nuclear reactor. A gas tight barrier or compartment of thermal insulation surrounds the selected components and includes devices to heat the internal atmosphere of the compartment. An external surface of the compartment or enclosure is cooled, such as by a circulating fluid. The heating devices provide for preheating of the components, as well as maintenance of a temperature sufficient to ensure that the reactor coolant fluid will not solidify during shutdown. The external cooling limits the heat transferred to other plant structures, such as supporting concrete and steel. The barrier is spaced far enough from the surrounded components so as to allow access for remote or manual inspection, maintenance, and repair.

  7. NEUTRONIC REACTOR OPERATIONAL METHOD AND CORE SYSTEM

    DOEpatents

    Winters, C.E.; Graham, C.B.; Culver, J.S.; Wilson, R.H.

    1960-07-19

    Homogeneous neutronic reactor systems are described wherein an aqueous fuel solution is continuously circulated through a spherical core tank. The pumped fuel solution-is injected tangentially into the hollow spherical interior, thereby maintaining vigorous rotation of the solution within the tank in the form of a vortex; gaseous radiolytic decomposition products concentrate within the axial vortex cavity. The evolved gas is continuously discharged through a gas- outlet port registering with an extremity of the vortex cavity. and the solution stream is discharged through an annular liquid outlet port concentrically encircling the gas outlet by virtue of which the vortex and its cavity are maintained precisely axially aligned with the gas outlet. A primary heat exchanger extracts useful heat from the hot effluent fuel solution before its recirculation into the core tank. Hollow cylinders and other alternative core- tank configurations defining geometric volumes of revolution about a principal axis are also covered. AEC's Homogeneous Reactor Experiment No. 1 is a preferred embodiment.

  8. Fuel handling system for a nuclear reactor

    DOEpatents

    Saiveau, James G.; Kann, William J.; Burelbach, James P.

    1986-01-01

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

  9. Fuel handling system for a nuclear reactor

    DOEpatents

    Saiveau, James G.; Kann, William J.; Burelbach, James P.

    1986-12-02

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

  10. The liquid annular reactor system (LARS) propulsion

    SciTech Connect

    Maise, G.; Lazareth, O.W.; Horn, F.; Powell, J.R.; Ludewig, H. ); Lenard, R.X. )

    1991-01-05

    A new concept for very high specific impulse ({gt}2000 seconds) direct nuclear propulsion is described. The concept, termed LARS (Liquid Annular Reactor System) uses liquid nuclear fuel elements to heat hydrogen propellant to very high temperatures ({similar to}6000 K). Operating pressure is moderate ({similar to}10 atm), with the result that the outlet hydrogen is virtually 100% dissociated to monatomic H. The molten fuel is contained in a solid container of its own material, which is rotated to stabilize the liquid layer by centripetal force. LARS reactor designs are described, together with neutronic and thermal-hydraulic analyses. Power levels are on the order of 200 megawatts. Typically, LARS designs use 7 rotating fuel elements, are beryllium moderated and have critical radii of {similar to}100 cm (core L/D{approx}1.5).

  11. Colliding Beam Fusion Reactor Space Propulsion System

    NASA Astrophysics Data System (ADS)

    Cheung, A.; Binderbauer, M.; Liu, F.; Qerushi, A.; Rostoker, N.; Wessel, F. J.

    2004-02-01

    The Colliding Beam Fusion Reactor Space Propulsion System, CBFR-SPS, is an aneutronic, magnetic-field-reversed configuration, fueled by an energetic-ion mixture of hydrogen and boron11 (H-B11). Particle confinement and transport in the CBFR-SPS are classical, hence the system is scaleable. Fusion products are helium ions, α-particles, expelled axially out of the system. α-particles flowing in one direction are decelerated and their energy recovered to ``power'' the system; particles expelled in the opposite direction provide thrust. Since the fusion products are charged particles, the system does not require the use of a massive-radiation shield. This paper describes a 100 MW CBFR-SPS design, including estimates for the propulsion-system parameters and masses. Specific emphasis is placed on the design of a closed-cycle, Brayton-heat engine, consisting of heat-exchangers, turbo-alternator, compressor, and finned radiators.

  12. Materials and mechanical design analysis of boron carbide reactor safety rods. Final report

    SciTech Connect

    Marra, J.C.

    1992-04-01

    The purpose of this task was to analyze the materials and mechanical design bases for the new boron carbide safety rod. These analyses included examination of the irradiation response of the materials, chemical compatibility of component materials, moisture considerations for the boron carbide pellets and susceptibility of the rod to corrosion under reactor environmental conditions. A number of issues concerning the mechanical behavior were also addressed. These included: safety rod dynamic response in scram scenarios, flexibility and mishandling behavior, and response to thermal excursions associated with gamma heating. A surveillance program aimed at evaluating the integrity of the safety rods following actual operating conditions and justifying life extension for the rods was also proposed. Based on the experimental testing and analyses associated with this task, it is concluded that the boron carbide safety rod design meets the materials and mechanical criteria for successful operational performance.

  13. Reflections on system safety and the law

    NASA Technical Reports Server (NTRS)

    Hayes, D. F., Sr.

    1971-01-01

    The application of law to the determination of what constitutes safeness is discussed. The numerous factors are analyzed which enter into the decisions of courts in deciding what is safe and what is unsafe. It is pointed out that as technology changes, legal interpretations of safety also change. Arguements are given for the use of system safety techniques and better engineering analyses as instruments of defense against liability.

  14. Reactor coolant pump monitoring and diagnostic system

    SciTech Connect

    Singer, R.M.; Gross, K.C.; Walsh, M. ); Humenik, K.E. )

    1990-01-01

    In order to reliably and safely operate a nuclear power plant, it is necessary to continuously monitor the performance of numerous subsystems to confirm that the plant state is within its prescribed limits. An important function of a properly designed monitoring system is the detection of incipient faults in all subsystems (with the avoidance of false alarms) coupled with an information system that provides the operators with fault diagnosis, prognosis of fault progression and recommended (either automatic or prescriptive) corrective action. In this paper, such a system is described that has been applied to reactor coolant pumps. This system includes a sensitive pattern-recognition technique based upon the sequential probability ratio test (SPRT) that detects incipient faults from validated signals, an expert system embodying knowledge bases on pump and sensor performance, extensive hypertext files containing operating and emergency procedures as well as pump and sensor information and a graphical interface providing the operator with easily perceived information on the location and character of the fault as well as recommended corrective action. This system is in the prototype stage and is currently being validated utilizing data from a liquid-metal cooled fast reactor (EBR-II). 3 refs., 4 figs.

  15. Safety assessment of high consequence robotics system

    SciTech Connect

    Robinson, D.G.; Atcitty, C.B.

    1996-08-01

    This paper outlines the use of a failure modes and effects analysis for the safety assessment of a robotic system being developed at Sandia National Laboratories. The robotic system, the weigh and leak check system, is to replace a manual process for weight and leakage of nuclear materials at the DOE Pantex facility. Failure modes and effects analyses were completed for the robotics process to ensure that safety goals for the systems have been met. Due to the flexible nature of the robot configuration, traditional failure modes and effects analysis (FMEA) were not applicable. In addition, the primary focus of safety assessments of robotics systems has been the protection of personnel in the immediate area. In this application, the safety analysis must account for the sensitivities of the payload as well as traditional issues. A unique variation on the classical FMEA was developed that permits an organized and quite effective tool to be used to assure that safety was adequately considered during the development of the robotic system. The fundamental aspects of the approach are outlined in the paper.

  16. The development of regulatory expectations for computer-based safety systems for the UK nuclear programme

    SciTech Connect

    Hughes, P. J.; Westwood, R.N; Mark, R. T.; Tapping, K.

    2006-07-01

    The Nuclear Installations Inspectorate (NII) of the UK's Health and Safety Executive (HSE) has completed a review of their Safety Assessment Principles (SAPs) for Nuclear Installations recently. During the period of the SAPs review in 2004-2005 the designers of future UK naval reactor plant were optioneering the control and protection systems that might be implemented. Because there was insufficient regulatory guidance available in the naval sector to support this activity the Defence Nuclear Safety Regulator (DNSR) invited the NII to collaborate with the production of a guidance document that provides clarity of regulatory expectations for the production of safety cases for computer based safety systems. A key part of producing regulatory expectations was identifying the relevant extant standards and sector guidance that reflect good practice. The three principal sources of such good practice were: IAEA Safety Guide NS-G-1.1 (Software for Computer Based Systems Important to Safety in Nuclear Power Plants), European Commission consensus document (Common Position of European Nuclear Regulators for the Licensing of Safety Critical Software for Nuclear Reactors) and IEC nuclear sector standards such as IEC60880. A common understanding has been achieved between the NII and DNSR and regulatory guidance developed which will be used by both NII and DNSR in the assessment of computer-based safety systems and in the further development of more detailed joint technical assessment guidance for both regulatory organisations. (authors)

  17. A thematic approach to system safety

    SciTech Connect

    Ekman, M.E.; Werner, P.W.; Covan, J.M.; D`Antonio, P.E.

    1997-12-01

    Sandia National Laboratories has refined a process for developing inherently safer system designs, based on methods used by the Laboratories to design detonation safety into nuclear weapons. The process was created when the Laboratories realized that standard engineering practices did not provide the level of safety assurance necessary for nuclear weapon operations, with their potential for catastrophic accidents. A systematic approach, which relies on mutually supportive design principles integrated through fundamental physical principles, was developed to ensure a predictably safe system response under a variety of operational and accident based stresses. Robust, safe system designs result from this thematic approach to safety, minimizing the number of safety critical features. This safety assurance process has two profound benefits: the process avoids the need to understand or limit the ultimate intensity of off normal environments and it avoids the requirement to analyze and test a bewildering and virtually infinite array of accident environment scenarios (e.g., directional threats, sequencing of environments, time races, etc.) to demonstrate conformance to all safety requirements.

  18. Progress in space nuclear reactor power systems technology development - The SP-100 program

    NASA Technical Reports Server (NTRS)

    Davis, H. S.

    1984-01-01

    Activities related to the development of high-temperature compact nuclear reactors for space applications had reached a comparatively high level in the U.S. during the mid-1950s and 1960s, although only one U.S. nuclear reactor-powered spacecraft was actually launched. After 1973, very little effort was devoted to space nuclear reactor and propulsion systems. In February 1983, significant activities toward the development of the technology for space nuclear reactor power systems were resumed with the SP-100 Program. Specific SP-100 Program objectives are partly related to the determination of the potential performance limits for space nuclear power systems in 100-kWe and 1- to 100-MW electrical classes. Attention is given to potential missions and applications, regimes of possible space power applicability, safety considerations, conceptual system designs, the establishment of technical feasibility, nuclear technology, materials technology, and prospects for the future.

  19. A guide for performing system safety analysis

    NASA Technical Reports Server (NTRS)

    Brush, J. M.; Douglass, R. W., III.; Williamson, F. R.; Dorman, M. C. (Editor)

    1974-01-01

    A general guide is presented for performing system safety analyses of hardware, software, operations and human elements of an aerospace program. The guide describes a progression of activities that can be effectively applied to identify hazards to personnel and equipment during all periods of system development. The general process of performing safety analyses is described; setting forth in a logical order the information and data requirements, the analytical steps, and the results. These analyses are the technical basis of a system safety program. Although the guidance established by this document cannot replace human experience and judgement, it does provide a methodical approach to the identification of hazards and evaluation of risks to the system.

  20. Manned space flight nuclear system safety. Volume 6: Space base nuclear system safety plan

    NASA Technical Reports Server (NTRS)

    1972-01-01

    A qualitative identification of the steps required to assure the incorporation of radiological system safety principles and objectives into all phases of a manned space base program are presented. Specific areas of emphasis include: (1) radiological program management, (2) nuclear system safety plan implementation, (3) impact on program, and (4) summary of the key operation and design guidelines and requirements. The plan clearly indicates the necessity of considering and implementing radiological system safety recommendations as early as possible in the development cycle to assure maximum safety and minimize the impact on design and mission plans.