WorldWideScience

Sample records for safety test facility reactor

  1. Study of fast reactor safety test facilities. Preliminary report

    International Nuclear Information System (INIS)

    Bell, G.I.; Boudreau, J.E.; McLaughlin, T.; Palmer, R.G.; Starkovich, V.; Stein, W.E.; Stevenson, M.G.; Yarnell, Y.L.

    1975-05-01

    Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods

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

    International Nuclear Information System (INIS)

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

    1992-07-01

    This report describes the EG ampersand 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

  3. Fast reactor test facilities in the US safety program

    International Nuclear Information System (INIS)

    Avery, R.; Dickerman, C.E.; Lennox, D.H.; Rose, D.

    1979-01-01

    The needs for safety information derivable from in-pile programs are reviewed, and the correlation made with existing and planned capability. In view of the current status of the U.S. breeder program, emphasis is given in the review to the impact of different fast breeder options on the required program and facilities. It is concluded that facility needs are somewhat independent of specific fast breeder concept, even though the relative emphasis on the various safety issues will differ. 8 refs

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

    International Nuclear Information System (INIS)

    Tomberlin, T.A.

    2002-01-01

    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. Physics design of fast reactor safety test facilities for in-pile experiments

    International Nuclear Information System (INIS)

    Travelli, A.; Matos, J.E.; Snelgrove, J.L.; Shaftman, D.H.; Tzanos, C.P.; Lam, S.K.; Pennington, E.M.; Woodruff, W.L.

    1976-01-01

    A determined effort to identify and resolve current Fast Breeder Reactor safety testing needs has recently resulted in a number of conceptual designs for FBR safety test facilities which are very complex and diverse both in their features and in their purpose. The paper discusses the physics foundations common to most fast reactor safety test facilities and the constraints which they impose on the design. The logical evolution, features, and capabilities of several major conceptual designs are discussed on the basis of this common background

  6. Safety report content and development for test loop facility on MARIA reactor

    International Nuclear Information System (INIS)

    Konechko, A.; Shumskij, A.M.; Mikul'ahin, V.E.

    1982-01-01

    A 600 kW test loop facility for investigatin.o safety problems is realized on MARIA reactor in Poland together with USSR organizations. Safety reports have been developed in two steps at the designstage. The 1st report being essentially a preliminary safety analysis was developed within the scope of the feasibility study. At the engineering design stage the preliminary test loop facility safety report had been prepared considering measures excluding the possibility of the MARIA reactor damage. The test loop facility safety report is fulfilled for normal, transient and emergency operation regimes. Separate safety basing for each group of experiments will be prepared. The report presents the test loop facility safety criteria coordinated by the nuclear safety comission. They contains the preliminary reports on the test loop facility safety. At the final stage of construction and at thecommitioning stage the start-up safety report will be developed which after required correction and adding up the putting into operation data will turn into operation safety report [ru

  7. An overview of FFTF [Fast Flux Test Facility] contributions to Liquid Metal Reactor Safety

    International Nuclear Information System (INIS)

    Waltar, A.E.; Padilla, A. Jr.

    1990-11-01

    The Fast Flux Test Facility has provided a very useful framework for testing the advances in Liquid Metal Reactor Safety Technology. During the licensing phase, the switch from a nonmechanistic bounding technique to the mechanistic approach was developed and implemented. During the operational phase, the consideration of new tests and core configurations led to use of the anticipated-transients-without-scram approach for beyond design basis events and the move towards passive safety. The future role of the Fast Flux Test Facility may involve additional passive safety and waste transmutation tests. 26 refs

  8. In-pile experiments and test facilities proposed for fast reactor safety

    International Nuclear Information System (INIS)

    Grolmes, M.A.; Avery, R.; Goldman, A.J.; Fauske, H.K.; Marchaterre, J.F.; Rose, D.; Wright, A.E.

    1976-01-01

    The role of in-pile experiments in support of the resolution of fast breeder reactor safety and licensing issues has been re-examined, with emphasis on key safety issues. Experiment needs have been related to the specific characteristics of these safety issues and to realistic requirements for additional test facility capabilities which can be achieved and utilized within the next ten years. It is found that those safety issues related to the energetics of core disruptive accidents have the largest impact on new facility requirements. However, utilization of existing facilities with modifications can provide for a continuing increase in experiment capability and experiment results on a timely bases. Emphasis has been placed upon maximum utilization of existing facilities and minimum requirements for new facilities. This evaluation has concluded that a new Safety Test Facility, STF, along with major modifications to the EBR II facility, improvement in TREAT capabilities, the existing Sodium Loop Safety Facility and corresponding Support Facilities provide the essential elements of the Safety Research Experiment Facilities (SAREF) required for resolution of key issues

  9. Relevance of passive safety testing at the fast flux test facility to advanced liquid metal reactors - 5127

    International Nuclear Information System (INIS)

    Wootan, D.W.; Omberg, R.P.

    2015-01-01

    Significant cost and safety improvements can be realized in advanced liquid metal reactor (LMR) designs by emphasizing inherent or passive safety through crediting the beneficial reactivity feedbacks associated with core and structural movement. This passive safety approach was adopted for the Fast Flux Test Facility (FFTF), and an experimental program was conducted to characterize the structural reactivity feedback. Testing at the Rapsodie and EBR-II reactors had demonstrated the beneficial effect of reactivity feedback caused by changes in fuel temperature and core geometry mechanisms in a liquid metal fast reactor in a holistic sense. The FFTF passive safety testing program was developed to examine how specific design elements influenced dynamic reactivity feedback in response to a reactivity input and to demonstrate the scalability of reactivity feedback results from smaller cores like Rapsodie and EBR-II to reactor cores that were more prototypic in scale to reactors of current interest. The U.S. Department of Energy, Office of Nuclear Energy Advanced Reactor Technology program is in the process of preserving, protecting, securing, and placing in electronic format information and data from the FFTF, including the core configurations and data collected during the passive safety tests. Evaluation of these actual test data could provide insight to improve analytical methods which may be used to support future licensing applications for LMRs. (authors)

  10. Physics constraints on the design of fast reactor safety test facilities

    International Nuclear Information System (INIS)

    Travelli, A.; Meneghetti, D.; Matos, J.; Snelgrove, J.; Shaftman, D.H.; Tzanos, C.; Lam, S.K.; Pennington, E.M.; Woodruff, W.L.

    1976-01-01

    This paper discusses the physics foundations common to all fast reactor safety test facilities and the constraints which they impose on the design. While detailed design discussions are confined to the experience with six ANL designs, available data from other designs are used to confirm the validity of the considerations and to broaden the scope of the discussion. This helps to view the various designs as a unified effort, to define their potential capabilities, and to assess how they could best complement each other

  11. The US Advanced Liquid Metal Reactor and the Fast Flux Test Facility Phase IIA passive safety tests

    International Nuclear Information System (INIS)

    Shen, P.K.; Harris, R.A.; Campbell, L.R.; Dautel, W.A.; Dubberley, A.E.; Gluekler, E.L.

    1992-07-01

    This report discusses the safety approach of the Advanced Liquid Metal reactor program, sponsored by the US Department of Energy, which relies upon passive reactor responses to off-normal condition to limit power and temperature excursions to levels that allow safety margins. Gas expansion modules (GEM) have included in the design to provide negative reactivity to enhance these margins in the extremely unlikely event that pumping power is lost and the highly reliable scram system fails to operate. The feasibility and beneficial features of these devices were first demonstrated in the core of the Fast Flux Test Facility (FFTF) in 1986. Preapplication safety evaluations by the US Nuclear Regulatory Commission have identified areas that must be addressed if these devices are to be relied on. One of these areas is the response of the reactor when it is critical and the pumps are turned on, resulting in positive reactivity being added to the core. Tests to examine such transients have been performed as part of the continuing FFTF program to confirm the passive safety characteristics of liquid metal reactors (LMR). The primary tests consisted of starting the main coolant pumps, which forced sodium coolant into the GEMS, decreasing neutron leakage and adding positive reactivity. The resulting transients were shown to be benign and easily mitigated by the reactivity feedbacks inherent in the FFTF and all LMRs. Steady-state auxiliary tests of the GEM and feedback reactivity worths accurately predicted the transient results. The auxiliary GEM worth tests also demonstrated that the worth can be determined at a subcritical state, which allows for a verification of the GEM's availability prior to ascending to power

  12. Review on conformance of JMTR reactor facility to safety design examination guides for water-cooled reactors for test and research

    International Nuclear Information System (INIS)

    Ide, Hiroshi; Naka, Michihiro; Sakuta, Yoshiyuki; Hori, Naohiko; Matsui, Yoshinori; Miyazawa, Masataka

    2009-03-01

    The safety design examination guides for water-cooled reactors for test and research are formulated as fundamental judgements on the basic design validity for licensing from a viewpoint of the safety. Taking the refurbishment opportunity of the JMTR, the conformance of the JMTR reactor facility to current safety design examination guides was reviewed with licensing documents, annexes and related documents. As a result, it was found that licensing documents fully satisfied the requirements of the current guides. Moreover, it was found that the JMTR reactor facility itself also satisfied the guides requirements as well as the safety performance, since the facility with safety function such as structure, systems, devices had been installed based on the licensing documents under the permission by the regulation authority. Important devices for safety have been produced under authorization of regulating authority. Therefore, it was confirmed that the licensing was conformed to guides, and that the JMTR has enough performance. (author)

  13. Ground test facility for nuclear testing of space reactor subsystems

    International Nuclear Information System (INIS)

    Quapp, W.J.; Watts, K.D.

    1985-01-01

    Two major reactor facilities at the INEL have been identified as easily adaptable for supporting the nuclear testing of the SP-100 reactor subsystem. They are the Engineering Test Reactor (ETR) and the Loss of Fluid Test Reactor (LOFT). In addition, there are machine shops, analytical laboratories, hot cells, and the supporting services (fire protection, safety, security, medical, waste management, etc.) necessary to conducting a nuclear test program. This paper presents the conceptual approach for modifying these reactor facilities for the ground engineering test facility for the SP-100 nuclear subsystem. 4 figs

  14. Study of In-Pile test facility for fast reactor safety research: performance requirements and design features

    Energy Technology Data Exchange (ETDEWEB)

    Nonaka, N.; Kawatta, N.; Niwa, H.; Kondo, S.; Maeda, K

    1996-12-31

    This paper describes a program and the main design features of a new in-pile safety facility SERAPH planned for future fast reactor safety research. The current status of R and D on technical developments in relation to the research objectives and performance requirements to the facility design is given.

  15. Irradiation Facilities at the Advanced Test Reactor

    International Nuclear Information System (INIS)

    S. Blaine Grover

    2005-01-01

    The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC) (formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950s with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world's data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens

  16. Advanced Test Reactor National Scientific User Facility

    International Nuclear Information System (INIS)

    Marshall, Frances M.; Benson, Jeff; Thelen, Mary Catherine

    2011-01-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  17. Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

    2011-08-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  18. New facilities in Japan materials testing reactor for irradiation test of fusion reactor components

    International Nuclear Information System (INIS)

    Kawamura, H.; Sagawa, H.; Ishitsuka, E.; Sakamoto, N.; Niiho, T.

    1996-01-01

    The testing and evaluation of fusion reactor components, i.e. blanket, plasma facing components (divertor, etc.) and vacuum vessel with neutron irradiation is required for the design of fusion reactor components. Therefore, four new test facilities were developed in the Japan Materials Testing Reactor: an in-pile functional testing facility, a neutron multiplication test facility, an electron beam facility, and a re-weldability facility. The paper describes these facilities

  19. Progress report concerning safety research for nuclear reactor facilities

    International Nuclear Information System (INIS)

    1978-01-01

    Examination and evaluation of safety research results for nuclear reactor facilities have been performed, as more than a year has elapsed since the plan had been initiated in April, 1976, by the special sub-committee for the safety of nuclear reactor facilities. The research is carried out by being divided roughly into 7 items, and seems to be steadily proceeding, though it does not yet reach the target. The above 7 items include researches for (1) criticality accident, (2) loss of coolant accident, (3) safety for light water reactor fuel, (4) construction safety for reactor facilities, (5) reduction of release of radioactive material, (6) safety evaluation based on the probability theory for reactor facilities, and (7) aseismatic measures for reactor facilities. With discussions on the progress and the results of the research this time, research on the behaviour on fuel in abnormal transients including in-core and out-core experiments has been added to the third item, deleting the power-cooling mismatch experiment in Nuclear Safety Research Reactor of JAERI. Also it has been decided to add two research to the seventh item, namely measured data collection, classification and analysis, and probability assessment of failures due to an earthquake. For these 7 items, the report describes the concrete contents of research to be performed in fiscal years of 1977 and 1978, by discussing on most rational and suitable contents conceivable at present. (Wakatsuki, Y.)

  20. Ventilation safety of facilities comprising nuclear reactors

    International Nuclear Information System (INIS)

    Guirlet, J.

    1982-01-01

    The reliability of the ventilation is one of the most important aspects in the prevention of the nuisances that a nuclear installation can provide, since the ventilation is located at the last barrier. A certain number of essential points have been recalled here. But it is necessary to bear in mind other requirements such as the limitation in the number of crossovers, the answers to be found should the system fail, the need to show that ventilation systems do not in themselves bring other nuisances such as noise, irradiation or contamination hazards, likelyhood of recycling the contamination, vibrations, fire. Finally, it is absolutely essential, right from the project stage, that the design ensures that very good accessibility, very easy dismantling and handling, as well as all the facilities needed to make sure of the initial and periodic tests, are guaranteed [fr

  1. Evaluation of neutronic characteristics of in-pile test reactor for fast reactor safety research

    Energy Technology Data Exchange (ETDEWEB)

    Uto, N.; Ohno, S.; Kawata, N. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1996-09-01

    An extensive research program has been carried out at the Power Reactor and Nuclear Fuel Development Corporation for the safety of future liquid-metal fast breeder reactors to be commercialized. A major part of this program is investigation and planning of advanced safety experiments conducted with a new in-pile safety test facility, which is larger and more advanced than any of the currently existing test reactors. Such a transient safety test reactor generally has unique neutronic characteristics that require various studies from the reactor physics point of view. In this paper, the outcome of the neutronics study is highlighted with presenting a reference core design concept and its performance in regard to the safety test objectives. (author)

  2. The selection of probabilistic safety assessment techniques for non-reactor nuclear facilities

    International Nuclear Information System (INIS)

    Vail, J.

    1992-01-01

    Historically, the probabilistic safety assessment (PSA) methodology of choice is the well known event tree/fault tree inductive technique. For reactor facilities is has stood the test of time. Some non-reactor nuclear facilities have found inductive methodologies difficult to apply. The stand-alone fault tree deductive technique has been used effectively to analyze risk in nuclear chemical processing facilities and waste handling facilities. The selection between the two choices suggest benefits from use of the deductive method for non-reactor facilities

  3. Passive safety testing at the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Lucoff, D.M.

    1989-01-01

    During 1986, the Fast Flux Test Facility (FFTF) conducted several tests designed to improve the understanding of the passive safety characteristics of an oxide-fueled liquid-metal reactor (LMR). Static and dynamic tests were performed over a broad range of power, flow, and temperature conditions that extended beyond those for normal operation. Key results of these tests are presented. Stable operation at low power with natural circulation cooling was demonstrated. A passive safety enhancement feature, the gas expansion module (GEM) was developed specifically to offset the large amount of cooldown reactivity that needs to be controlled in an oxide-fueled LMR undergoing an unprotected loss-of-flow accident. Nine GEMs were built and successfully tested in FFTF. With the reactor at 50% power (200 MW (thermal)), the main coolant pumps were turned off and the normal control rod scram response was inhibited. The GEMs and inherent core reactivity feedback mechanisms took the core subcritical with a modest peak coolant temperature transient that reached 85 degrees C above the pretransient value and always maintained a >400 degrees C margin to the sodium boiling point (910 degrees C)

  4. Experimental facilities for gas-cooled reactor safety studies. Task group on Advanced Reactor Experimental Facilities (TAREF)

    International Nuclear Information System (INIS)

    2009-01-01

    In 2007, the NEA Committee on the Safety of Nuclear Installations (CSNI) completed a study on Nuclear Safety Research in OECD Countries: Support Facilities for Existing and Advanced Reactors (SFEAR) which focused on facilities suitable for current and advanced water reactor systems. In a subsequent collective opinion on the subject, the CSNI recommended to conduct a similar exercise for Generation IV reactor designs, aiming to develop a strategy for ' better preparing the CSNI to play a role in the planned extension of safety research beyond the needs set by current operating reactors'. In that context, the CSNI established the Task Group on Advanced Reactor Experimental Facilities (TAREF) in 2008 with the objective of providing an overview of facilities suitable for performing safety research relevant to gas-cooled reactors and sodium fast reactors. This report addresses gas-cooled reactors; a similar report covering sodium fast reactors is under preparation. The findings of the TAREF are expected to trigger internationally funded CSNI projects on relevant safety issues at the key facilities identified. Such CSNI-sponsored projects constitute a means for efficiently obtaining the necessary data through internationally co-ordinated research. This report provides an overview of experimental facilities that can be used to carry out nuclear safety research for gas-cooled reactors and identifies priorities for organizing international co-operative programmes at selected facilities. The information has been collected and analysed by a Task Group on Advanced Reactor Experimental Facilities (TAREF) as part of an ongoing initiative of the NEA Committee on the Safety of Nuclear Installations (CSNI) which aims to define and to implement a strategy for the efficient utilisation of facilities and resources for Generation IV reactor systems. (author)

  5. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Directory of Open Access Journals (Sweden)

    Hwang Bae

    2017-08-01

    Full Text Available Three small-break loss-of-coolant accident (SBLOCA tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor, i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  6. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Hwang; Ryu, Sung Uk; Yi, Sung Jae; Park, Hyun Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Dong Eok [Dept. of Precision Mechanical Engineering, Kyungpook National University, Sangju (Korea, Republic of)

    2017-08-15

    Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  7. Thermal-hydraulic tests for reactor safety system

    International Nuclear Information System (INIS)

    Chun, Se Young; Chung, Moon Ki; Baek, Won Pil

    2002-05-01

    Tests for the safety depressurization system, Sparger adopted for the Korean next generation reactor, APR1400 are carried out for several geometries with the B and C (Blowdown and Condensation) facility in the condition of high temperature and pressure and with a small test facility in the condition of atmospheric temperature and pressure. Tests for the critical heat flux are performed with the RCS(Reactor Coolant System) facility as well as with the Freon CHF Loop in the condition of high temperature and pressure. The atmospheric temperature and pressure facility is utilized for development of the high standard thermal hydraulic measurement technology. The optical method is developed to measure the local thermal-hydraulic behavior for the single and two-phase boiling phenomena

  8. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    OpenAIRE

    Hwang Bae; Dong Eok Kim; Sung-Uk Ryu; Sung-Jae Yi; Hyun-Sik Park

    2017-01-01

    Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are s...

  9. Dynamic Response Testing in an Electrically Heated Reactor Test Facility

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Morton, T. J.

    2006-01-01

    Non-nuclear testing can be a valuable tool in development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system (Bragg-Sitton, 2005). The current paper applies the same testing methodology to a direct drive gas cooled reactor system, demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. In each testing application, core power transients were controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. Although both system designs utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    T. A. Tomberlin; S. B. Grover

    2004-01-01

    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

  12. Proposal for a seismic facility for reactor safety research

    International Nuclear Information System (INIS)

    Anderson, C.A.; Dove, R.C.; Rhorer, R.L.

    1976-07-01

    Certain problem areas in the seismic analysis and design of nuclear reactors are enumerated and the way in which an experimental program might contribute to each area is examined. The use of seismic simulation testing receives particular attention, especially with regard to the verification of structural response analysis. The importance of scale modeling used in conjunction with seismic simulation is also stressed. The capabilities of existing seismic simulators are summarized, and a proposed facility is described which would considerably extend the ability to conduct, with confidence, confirmatory experiments on the behavior of reactor components when subjected to seismic excitation. Particular applications to gas-cooled and other reactor types are described

  13. Thermal Hydraulic Tests for Reactor Core Safety

    Energy Technology Data Exchange (ETDEWEB)

    Moon, S. K.; Baek, W. P.; Chun, S. Y. (and others)

    2007-06-15

    The main objectives of the present project are to resolve the current issues of reactor core thermal hydraulics, to develop an advanced measurement and analytical techniques, and to perform reactor core safety verification tests. 6x6 reflood experiments, various heat transfer experiments using Freon, and experiments on the spacer grids effects on the post-dryout are carried out using spacer grids developed in Korea in order to resolve the current issues of the reactor core thermal hydraulics. In order to develop a reflood heat transfer model, the detailed reflood phenomena are visualized and measured using round tube and 2x2 rod bundle. A detailed turbulent mixing phenomenon for subchannels is measured using advanced measurement techniques such as LDV and PIV. MARS and MATRA codes developed in Korea are assessed, verified and improved using the obtained experimental data. Finally, a systematic quality assurance program and experimental data generation system has been constructed in order to increase the reliability of the experimental data.

  14. Safety Research Experiment Facility Project. Conceptual design report. Volume VII. Reactor cooling

    International Nuclear Information System (INIS)

    1975-12-01

    The Reactor Cooling System (RCS) will provide the required cooling during test operations of the Safety Research Experiment Facility (SAREF) reactor. The RCS transfers the reactor energy generated in the core to a closed-loop water storage system located completely inside the reactor containment building. After the reactor core has cooled to a safe level, the stored heat is rejected through intermediate heat exchangers to a common forced-draft evaporative cooling tower. The RCS is comprised of three independent cooling loops of which any two can remove sufficient heat from the core to prevent structural damage to the system components

  15. Tests for validation of fast neutron reactors safety

    International Nuclear Information System (INIS)

    Nagata, T.; Yamashita, H.

    2001-01-01

    Japanese scientific research and design enterprises in cooperation with industrial and power generating corporations implement a project on creating a fast neutron reactor of the ultimate safety. One of the basic expected results from such a development is creation of a reactor core structure that is able to eliminate recriticality occurrence in the course of reactor accident involving fuel melting. One of the possible ways to solve this problem is to include pipes (meant for specifying directed (controlled) molten fuel relocation) into fuel assembly structure. In the course of conduction and subsequent implementation of such a design the basic issue is to experimentally confirm the operating capacity of FA having such a structure and that is called FAIDUS. Within EAGLE Project on experimental basis of IAE NNC RK an activity has been started on preparation and conduction of out-of-pile and in-pile tests. During tests a sodium coolant will be used. Studies are conducted by NNC RK in cooperation with the Japanese corporations JAPC and JNC. Basic objective of out-of-pile tests was to obtain preliminary information on fuel relocation behavior under conditions simulating accident involving melting of core consisting of FAIDUS FA, which will help to clarify simulation criteria and to develop the most optimum structure of the experimental channel for reactor experiments conduction. The basic objective of in-pile tests was the experimental confirmation of operating capacity of FAIDUS FA model under reactor conditions. According to the program two tests are planned to be performed at IGR reactor: tests for validation of fast neutron reactor safety, and out-of-pile tests at EAGLE experimental facility without sodium coolant

  16. Safety evaluation report. Fast Flux Test Facility. Project No. 448

    Energy Technology Data Exchange (ETDEWEB)

    1978-08-01

    Information on the safety of the FFTF Reactor is presented under the following chapter headings: site characteristics; design of structures, components, equipment, and systems; reactor; reactor coolant system and connected systems; engineered safety features; electric power; auxiliary systems; radioactive waste management systems; radiation protection; conduct of operations; initial test programs; accident analysis; and quality assurance.

  17. Safety evaluation report. Fast Flux Test Facility. Project No. 448

    International Nuclear Information System (INIS)

    1978-01-01

    Information on the safety of the FFTF Reactor is presented under the following chapter headings: site characteristics; design of structures, components, equipment, and systems; reactor; reactor coolant system and connected systems; engineered safety features; electric power; auxiliary systems; radioactive waste management systems; radiation protection; conduct of operations; initial test programs; accident analysis; and quality assurance

  18. Safety problems encountered in construction and operation of the sodium test facilities of the Institute of Reactor Development (IRD) at the Karlsruhe Nuclear Research Center

    International Nuclear Information System (INIS)

    Schleisiek, K.

    1971-01-01

    In this report the safety aspects of the design and construction of a sodium boiling loop and a sodium tank test facility are discussed. Subsequently two experiments concerning the safety of the facilities are described: the testing of a drip basin to collect the sodium and to limit the rate of burning in the case of a leak, and the investigation of the chemical reaction of sodium with the insulating materials. Finally some general emergency procedures in the case of sodium incidents are discussed. A 16 mm-film demonstrating sodium fires and fire fighting methods will be shown. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

  1. A review of experiments and results from the transient reactor test (TREAT) facility

    International Nuclear Information System (INIS)

    Deitrich, L. W.

    1998-01-01

    The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop

  2. Safety test facilities - status, needs, future directions

    International Nuclear Information System (INIS)

    Heusener, G.; Cogne, F.

    1979-08-01

    A survey is given of the in-pile programs which are presently or in the near future being performed in the DeBeNe-area and in France. Only those in-pile programs are considered which are dealing with severe accidents that might lead to disruption of major parts of the core. By comparing the needs with the goals of the present programs points are identified which are not sufficiently well covered up till now. The future procedure is described: the existing facilities will be used to the largest possible extent. Whenever it is necessary, upgrading and improvement will be foreseen. Studies of a Test Facility allowing the transient testing of large pin bundles should be continued. The construction of such a facility in Europe in the near future however seems premature

  3. Advanced Test Reactor National Scientific User Facility Partnerships

    International Nuclear Information System (INIS)

    Marshall, Frances M.; Allen, Todd R.; Benson, Jeff B.; Cole, James I.; Thelen, Mary Catherine

    2012-01-01

    In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of Wisconsin

  4. Safety assessment for the rf Test Facility

    International Nuclear Information System (INIS)

    Nagy, A.; Beane, F.

    1984-08-01

    The Radio Frequency Test Facility (RFTF) is a part of the Magnetic Fusion Program's rf Heating Experiments. The goal of the Magnetic Fusion Program (MFP) is to develop and demonstrate the practical application of fusion. RFTF is an experimental device which will provide an essential link in the research effort aiming at the realization of fusion power. This report was compiled as a summary of the analysis done to ensure the safe operation of RFTF

  5. Advanced reactor experimental facilities

    International Nuclear Information System (INIS)

    Amri, A.; Papin, J.; Uhle, J.; Vitanza, C.

    2010-01-01

    For many years, the NEA has been examining advanced reactor issues and disseminating information of use to regulators, designers and researchers on safety issues and research needed. Following the recommendation of participants at an NEA workshop, a Task Group on Advanced Reactor Experimental Facilities (TAREF) was initiated with the aim of providing an overview of facilities suitable for carrying out the safety research considered necessary for gas-cooled reactors (GCRs) and sodium fast reactors (SFRs), with other reactor systems possibly being considered in a subsequent phase. The TAREF was thus created in 2008 with the following participating countries: Canada, the Czech Republic, Finland, France, Germany, Hungary, Italy, Japan, Korea and the United States. In a second stage, India provided valuable information on its experimental facilities related to SFR safety research. The study method adopted entailed first identifying high-priority safety issues that require research and then categorizing the available facilities in terms of their ability to address the safety issues. For each of the technical areas, the task members agreed on a set of safety issues requiring research and established a ranking with regard to safety relevance (high, medium, low) and the status of knowledge based on the following scale relative to full knowledge: high (100%-75%), medium (75 - 25%) and low (25-0%). Only the issues identified as being of high safety relevance and for which the state of knowledge is low or medium were included in the discussion, as these issues would likely warrant further study. For each of the safety issues, the TAREF members identified appropriate facilities, providing relevant information such as operating conditions (in- or out-of reactor), operating range, description of the test section, type of testing, instrumentation, current status and availability, and uniqueness. Based on the information collected, the task members assessed prospects and priorities

  6. Safety re-assessment of AECL test and research reactors

    International Nuclear Information System (INIS)

    Winfield, D.J.

    1990-01-01

    Atomic Energy of Canada Limited currently has four operating engineering test/research reactors of various sizes and ages; a new isotope-production reactor Maple-X10, under construction at Chalk River Nuclear Laboratories (CRNL), and a heating demonstration reactor, SDR, undergoing high-power commissioning at Whiteshell Nuclear Research Establishment (WNRE). The company is also performing design studies of small reactors for hot water and electricity production. The older reactors are ZED-2, PTR, NRX, and NRU; these range in age from 42 years (NRX) to 29 years (ZED-2). Since 1984, limited-scope safety re-assessments have been underway on three of these reactors (ZED-2, NRX AND NRU). ZED-2 and PTR are operated by the Reactor Physics Branch; all other reactors are operated by the respective site Reactor Operations Branches. For the older reactors the original safety reports produced were entirely deterministic in nature and based on the design-basis accident concept. The limited scope safety re-assessments for these older reactors, carried out over the past 5 years, have comprised both quantitative probabilistic safety-assessment techniques, such as event tree and fault analysis, and/or qualitative techniques, such as failure mode and effect analysis. The technique used for an individual assessment was dependent upon the specific scope required. This paper discusses the types of analyses carried out, specific insights/recommendations resulting from the analysis, and the plan for future analysis. In addition, during the last four years safety assessments have been carried out on the new isotope-, heat-, and electricity-producing reactors, as part of the safety design review, commissioning and licensing activities

  7. Safety significance of ATR [Advanced Test Reactor] passive safety response attributes

    International Nuclear Information System (INIS)

    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

  8. Facility for in-reactor creep testing of fuel cladding

    International Nuclear Information System (INIS)

    Kohn, E.; Wright, M.G.

    1976-11-01

    A biaxial stress creep test facility has been designed and developed for operation in the WR-1 reactor. This report outlines the rationale for its design and describes its construction and the operating experience with it. The equipment is optimized for the determination of creep data on CANDU fuel cladding. Typical results from Zr-2.5 wt% Nb fuel cladding are used to illustrate the accuracy and reliability obtained. (author)

  9. The role of natural circulation in the FFTF [Fast Flux Test Facility] passive safety tests

    International Nuclear Information System (INIS)

    Stover, R.L.; Padilla, A.; Burke, T.M.; Knecht, W.L.

    1987-03-01

    A series of tests were completed at the Fast Flux Test Facility to demonstrate the passive safety characteristics of liquid metal reactors with natural circulation flow. The first test consisted of transition from forced to natural circulation flow at an initial decay power of 0.3%. The second test represented an unprotected loss-of-flow transient to natural circulation from 50% power with the control rods prevented from scramming into the core. The third test was a steady-state, natural circulation condition with core fission powers up ato about 2.3%. Core sodium data and results of single and multi-channel computer models confirmed the reliability and effectiveness of natural circulation flow for liquid metal reactor safety

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

    International Nuclear Information System (INIS)

    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

  11. Fast reactor safety testing in Transient Reactor Test (TREAT) in the 1980s

    International Nuclear Information System (INIS)

    Wright, A.E.; Dutt, D.S.; Harrison, L.J.

    1990-01-01

    Several series of fast reactor safety tests were performed in TREAT during the 1980s. These focused on the transient behavior of full-length oxide fuels (US reference, UK reference, and US advanced design) and on modern metallic fuels. Most of the tests addressed fuel behavior under transient overpower or loss-of-flow conditions. The test series were the PFR/TREAT tests; the RFT, TS, CDT, and RX series on oxide fuels; and the M series on metallic fuels. These are described in terms of their principal results and relevance to analyses and safety evaluation. 4 refs., 3 tabs

  12. The Advanced Test Reactor Irradiation Facilities and Capabilities

    International Nuclear Information System (INIS)

    S. Blaine Grover; Raymond V. Furstenau

    2007-01-01

    The Advanced Test Reactor (ATR) is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. The ATR is a very versatile facility with a wide variety of experimental test capabilities for providing the environment needed in an irradiation experiment. These different capabilities include passive sealed capsule experiments, instrumented and/or temperature-controlled experiments, and pressurized water loop experiment facilities. The ATR has enhanced capabilities in experiment monitoring and control systems for instrumented and/or temperature controlled experiments. The control systems utilize feedback from thermocouples in the experiment to provide a custom blended flowing inert gas mixture to control the temperature in the experiments. Monitoring systems have also been utilized on the exhaust gas lines from the experiment to monitor different parameters, such as fission gases for fuel experiments, during irradiation. ATR's unique control system provides axial flux profiles in the experiments, unperturbed by axially positioned control components, throughout each reactor operating cycle and over the duration of test programs requiring many years of irradiation. The ATR irradiation positions vary in diameter from 1.6 cm (0.625 inches) to 12.7 cm (5.0 inches) over an active core length of 122 cm (48.0 inches). Thermal and fast neutron fluxes can be adjusted radially across the core depending on the needs of individual test programs. This paper will discuss the different irradiation capabilities available and the cost/benefit issues related to each capability. Examples of different experiments will also be discussed to demonstrate the use of the capabilities and facilities at ATR for performing irradiation experiments

  13. Fast Flux Test Facility final safety analysis report. Amendment 72

    Energy Technology Data Exchange (ETDEWEB)

    Gantt, D. A.

    1992-08-01

    This document provides the Final Safety Analysis Report (FSAR) Amendment 72 for incorporation into the Fast Flux Test Facility (FFTF) FSAR set. This amendment change incorporates Engineering Change Notices issued subsequent to Amendment 71 and approved for incorporation before June 24, 1992. These include changes in: Chapter 2, Site Characteristics; Chapter 3, Design Criteria Structures, Equipment, and Systems; Chapter 5B, Reactor Coolant System; Chapter 7, Instrumentation and Control Systems; Chapter 8, Electrical Systems - The description of the Class 1E, 125 Vdc systems is updated for the higher capacity of the newly installed, replacement batteries; Chapter 9, Auxiliary Systems - The description of the inert cell NASA systems is corrected to list the correct number of spare sample points; Chapter 11, Reactor Refueling System; Chapter 12, Radiation Protection and Waste Management; Chapter 13, Conduct of Operations; Chapter 16, Quality Assurance; Chapter 17, Technical Specifications; Chapter 19, FFTF Fire Specifications for Fire Detection, Alarm, and Protection Systems; Chapter 20, FFTF Criticality Specifications; and Appendix B, Primary Piping Integrity Evaluation.

  14. RIA testing capability of the transient reactor test facility

    International Nuclear Information System (INIS)

    Crawford, D.C.; Swanson, R.W.

    1999-01-01

    The advent of high-burnup fuel implementation in LWRs has generated international interest in high-burnup LWR fuel performance. Recent testing under simulated RIA conditions has demonstrated that certain fuel designs fail at peak fuel enthalpy values that are below existing regulatory criteria. Because many of these tests were performed with non-prototypically aggressive test conditions (i.e., with power pulse widths less than 10 msec FWHM and with non-protoypic coolant configurations), the results (although very informative) do not indisputably identify failure thresholds and fuel behavior. The capability of the TREAT facility to perform simulated RIA tests with prototypic test conditions is currently being evaluated by ANL personnel. TREAT was designed to accommodate test loops and vehicles installed for in-pile transient testing. During 40 years of TREAT operation and fuel testing and evaluation, experimenters have been able to demonstrate and determine the transient behavior of several types of fuel under a variety of test conditions. This experience led to an evolution of test methodology and techniques which can be employed to assess RIA behavior of LWR fuel. A pressurized water loop that will accommodate RIA testing of LWR and CANDU-type fuel has completed conceptual design. Preliminary calculations of transient characteristics and energy deposition into test rods during hypothetical TREAT RIA tests indicate that with the installation of a pressurized water loop, the facility is quite capable of performing prototypic RIA testing. Typical test scenarios indicate that a simulated RIA with a 72 msec FWHM pulse width and energy deposition of 1200 kJ/kg (290 cal/gm) is possible. Further control system enhancements would expand the capability to pulse widths as narrow as 40 msec. (author)

  15. Gas reactor in-pile safety test project (GRIST-2)

    International Nuclear Information System (INIS)

    Kelley, A.P. Jr.; Arbtin, E.; St Pierre, R.

    1979-01-01

    Although out-of-pile tests may be expected to confirm individual phenomena models in core disruptive accident analysis codes, only in-pile tests are capable of verifying the extremely complex integrated model effects within the appropriate time phase for these accidents. For this reason, the GRIST-2 project, the purpose of which is to design and construct an in-pile helium loop capable of transient safety testing in the TREAT facility in Idaho, forms a cornerstone of the US GCFR safety program. The project organization, experiment program, facility, helium system design, and schedule which have been selected to meet the objectives are described

  16. Procedures for conducting probabilistic safety assessment for non-reactor nuclear facilities

    International Nuclear Information System (INIS)

    2002-01-01

    A well performed and adequately documented safety assessment of a nuclear facility will serve as a basis to determine whether the facility complies with the safety objectives, principles and criteria as stipulated by the national regulatory body of the country where the facility is in operation. International experience shows that the practices and methodologies used to perform safety assessments and periodic safety re-assessment for non-reactor nuclear facilities differ significantly from county to country. Most developing countries do not have methods and guidance for safety assessment that are prescribed by the regulatory body. Typically the safety evaluation for the facility is based on a case by case assessment. Whilst conservative deterministic analyses are predominantly used as a licensing basis in many countries, recently probabilistic safety assessment (PSA) techniques have been applied as a useful complementary tool to support safety decision making. The main benefit of PSA is to provide insights into the safety aspects of facility design and operation. PSA points up the potential environmental impacts of postulated accidents, including the dominant risk contributors, and enables safety analysts to compare options for reducing risk. In order to advise on how to apply PSA methodology for the safety assessment of non-reactor nuclear facilities, the IAEA organized several consultants meetings, which led to the preparation of this TECDOC. This document is intended as guidance for the conduct of PSA in non-nuclear facilities. The main emphasis here is on the general procedural steps of a PSA that is specific for a non-reactor nuclear facility, rather than the details of the specific methods. The report is directed at technical staff managing or performing such probabilistic assessments and to promote a standardized framework, terminology and form of documentation for these PSAs. It is understood that the level of detail implied in the tasks presented in this

  17. PANDA a multi-purpose thermal-hydraulics facility devoted to nuclear reactor containment safety analysis

    International Nuclear Information System (INIS)

    Paladino, Domenico

    2014-01-01

    This paper presents the multi purpose facility PANDA devised for the safety analysis of nuclear reactor containment. The passive safety systems for LWRs have been explained with details about the PAssive Nachzerfallswärmeabfuhr und Druck-Abbau Testanlage (PANDA)

  18. PANDA: A Multipurpose Integral Test Facility for LWR Safety Investigations

    International Nuclear Information System (INIS)

    Paladino, D.; Dreier, J.

    2012-01-01

    The PANDA facility is a large scale, multicompartmental thermal hydraulic facility suited for investigations related to the safety of current and advanced LWRs. The facility is multipurpose, and the applications cover integral containment response tests, component tests, primary system tests, and separate effect tests. Experimental investigations carried on in the PANDA facility have been embedded in international projects, most of which under the auspices of the EU and OECD and with the support of a large number of organizations (regulatory bodies, technical dupport organizations, national laboratories, electric utilities, industries) worldwide. The paper provides an overview of the research programs performed in the PANDA facility in relation to BWR containment systems and those planned for PWR containment systems.

  19. French safety authority projects in the field of research and test reactors

    International Nuclear Information System (INIS)

    Saint Raymond, P.; Duthe, M.; Abou Yehia, H.

    2001-01-01

    This paper gives an outline of some actions initiated by the French safety authority in the field of research and test reactors. An important action concerns the definition of the authorisation criteria for the implementation of experiments in these reactors. In particular, it is necessary to define clearly in which conditions an experiment may be authorised internally by the operating organisation or needs a formal approval by the safety authority. The practice related to the systematic safety reassessment of old facilities and the regulatory provisions associated with the decommissioning are presented after a discussion on the ageing issues. (author)

  20. Reactor cover gas monitoring at the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Bechtold, R.A.; Holt, F.E.; Meadows, G.E.; Schenter, R.E.

    1986-09-01

    The Fast Flux Test Facility (FFTF) is a 400-megawatt (thermal) sodium-cooled reactor designed for irradiation testing of fuels, materials and components for LMRs. It is operated by the Westinghouse Hanford Company for the US Department of Energy on the government-owned Hanford reservation near Richland, Washington. The first 100-day operating cycle began in April 1982 and the eighth operating cycle was completed in July 1986. Argon is used as the cover gas for all sodium systems at the plant. A program for cover gas monitoring has been in effect since the start of sodium fill in 1978. The argon is supplied to the FFTF by a liquid argon Dewar System and used without further purification

  1. An independent safety assessment of Department of Energy nuclear reactor facilities: Safety overview and management function

    International Nuclear Information System (INIS)

    Booth, M.; Brodsky, R.S.; Frankhouser, W.L.

    1981-02-01

    The Under Secretary of Energy established the Nuclear Facilities Personnel Qualification and Training (NFPQT) Committee in October, 1979, in the aftermath of the Three Mile Island (TMI) nuclear accident, to assess the adequacy of training of personnel at DOE nuclear facilities. Subsequently, in February, 1980, the charge to this Committee was modified to assess all implications of the Kemeny Commission report on TMI with regard to DOE nuclear reactors, excluding those in the Division of Naval Reactors. The modified charge was also limited, for the time being, to reactor facilities instead of all nuclear facilities. This report describes the portion of the revised assessment activities that was assigned to the Assessment Support Team

  2. Safety analysis calculations for research and test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chen, S Y; MacDonald, R; MacFarlane, D [Argonne National Laboratory, Argonne, IL (United States)

    1983-08-01

    The goal of the RERTR (Reduced Enrichment in Research and Test Reactor) Program at ANL is to provide technical means for conversion of research and test reactors from HEU (High-Enrichment Uranium) to LEU (Low-Enrichment Uranium) fuels. In exploring the feasibility of conversion, safety considerations are a prime concern; therefore, safety analyses must be performed for reactors undergoing the conversion. This requires thorough knowledge of the important safety parameters for different types of reactors for both HEU and LEU fuel. Appropriate computer codes are needed to predict transient reactor behavior under postulated accident conditions. In this discussion, safety issues for the two general types of reactors i.e., the plate-type (MTR-type) reactor and the rod-type (TRIGA-type) reactor, resulting from the changes associated with LEU vs. HEU fuels, are explored. The plate-type fuels are typically uranium aluminide (UAl{sub x}) compounds dispersed in aluminum and clad with aluminum. Moderation is provided by the water coolant. Self shut-down reactivity coefficients with EU fuel are entirely a result of coolant heating, whereas with LEU fuel there is an additional shut down contribution provided by the direct heating of the fuel due to the Doppler coefficient. In contrast, the rod-type (TRIGA) fuels are mixtures of zirconium hydride, uranium, and erbium. This fuel mixture is formed into rods ( {approx} 1 cm diameter) and clad with stainless steel or Incoloy. In the TRIGA fuel the self-shutdown reactivity is more complex, depending on heating of the fuel rather than the coolant. The two most important mechanisms in providing this feedback are: spectral hardening due to neutron interaction with the ZrH moderator as it is heated and Doppler broadening of resonances in erbium and U-238. Since these phenomena result directly from heating of the fuel, and do not depend on heat transfer to the moderator/coolant, the coefficients are prompt acting. Results of transient

  3. Safety evaluation of the loss of fluid test facility project No. 394

    International Nuclear Information System (INIS)

    1975-05-01

    Assessment of the safety of the LOFT facility and subsequent recommendations have been based on a comparison of the LOFT facility to requirements for commercial power reactors. In this comparison, the many unique features of the LOFT facility were considered including the low power level, the limited operational use as a test reactor, and the remoteness of the site. Based on this assessment, it is concluded, that while the likelihood of an accidental release of fission products may be greater than for a commercial power reactor, the consequences of such a release are reduced by the lower fission product inventory, the remoteness of the site and the capability of evacuating the Idaho National Engineering Laboratory (INEL) and adjacent areas. There is reasonable assurance that the public health and safety will not be endangered due to operation of this facility, specifically: The INEL site is acceptable with respect to location, land use, population distribution, controlled access, hydrology, meteorology, geology and seismology. Sufficient engineered safety features have been included to assure that the potential offsite doses are well within 10 CFR Part 100 guidelines. The LOFT facility has been designed in general accordance with standards, guides and codes which are comparable to those applied to commercial power reactors and any exceptions to these have been based on the unique features of the LOFT facility. Certain matters including the final safety analyses based on detailed component designs, Technical Specifications, LOCE controls and detailed program plan have not been reviewed but we assume will properly be resolved by ERDA, which has the ultimate responsibility for the safety of this facility. Changes to the facility design or program plan such as removal of the fueled Mobile Test Assembly or blowdowns to the containment vessel also will require additional analyses and review. (U.S.)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    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.

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

    International Nuclear Information System (INIS)

    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

  6. Integral test facilities for validation of the performance of passive safety systems and natural circulation

    International Nuclear Information System (INIS)

    Choi, J. H.

    2010-10-01

    Passive safety systems are becoming an important component in advanced reactor designs. This has led to an international interest in examining natural circulation phenomena as this may play an important role in the operation of these passive safety systems. Understanding reactor system behaviour is a challenging process due to the complex interactions between components and associated phenomena. Properly scaled integral test facilities can be used to explore these complex interactions. In addition, system analysis computer codes can be used as predictive tools in understanding the complex reactor system behaviour. However, before the application of system analysis computer codes for reactor design, it is capability in making predictions needs to be validated against the experimental data from a properly scaled integral test facility. The IAEA has organized a coordinated research project (CRP) on natural circulation phenomena, modeling and reliability of passive systems that utilize natural circulation. This paper is a part of research results from this CRP and describes representative international integral test facilities that can be used for data collection for reactor types in which natural circulation may play an important role. Example experiments were described along with the analyses of these example cases in order to examine the ability of system codes to model the phenomena that are occurring in the test facilities. (Author)

  7. Evaluation of safety test needs for the gas cooled breeder reactors

    International Nuclear Information System (INIS)

    Emon, D.E.; Buttemer, D.R.; Sevy, R.H.

    1976-01-01

    This paper deals with the process used in determining the safety test needs for the Gas Cooled Fast Breeder Reactor (GCFR), reports existing tentative conclusions, and indicates the direction that the process is taking at this time. The process is based upon two ideas: (1) that the safety information needs will be identified through risk analysis directly dependent on the various design features of the GCFR and (2) that the safety program will be determined by a safety review committee. The paper limits itself to presenting thoughts on the safety test needs directly associated with the GCFR core during severe beyond design basis accident situations involving the loss of coolable core geometry. Representative event sequence diagrams are reported for the three generic classes of accidents considered. The following categories of information are identified: safety information needs, safety tests required to fulfill these information needs, and the facilities required to perform the tests

  8. Reactor safety

    International Nuclear Information System (INIS)

    Butz, H.P.; Heuser, F.W.; May, H.

    1985-01-01

    The paper comprises an introduction into nuclear physics bases, the safety concept generally speaking, safety devices of pwr type reactors, accident analysis, external influences, probabilistic safety assessment and risk studies. It further describes operational experience, licensing procedures under the Atomic Energy Law, research in reactor safety and the nuclear fuel cycle. (DG) [de

  9. The roles of EBR-II and TREAT [Transient Reactor Test] in establishing liquid metal reactor safety

    International Nuclear Information System (INIS)

    Sackett, J.I.; Lehto, W.K.; Solbrig, C.W.

    1990-01-01

    This paper examines the role of the Experimental Breeder Reactor II (EBR-II) and Transient Reactor Test (TREAT) facilities in contributing to the understanding and resolution of key safety issues in liquid metal reactor safety during the decade of the 80's. Fuels and materials testing has been carried out to address questions on fuels behavior during steady-state and upset conditions. In addition, EBR-II has conducted plant tests to demonstrate passive response to ATWS events and to develop control and diagnostic strategies for safe operation of advanced LMRs. TREAT and EBR-II complement each other and between them provide a transient testing capability that covers the whole range of concerns during overpower conditions. EBR-II, with use of the special Automatic Control Rod Drive System, can generate power change rates that overlap the lower end of the TREAT capability. 21 refs

  10. Design of Safety Parameter Monitoring Function in a Research Reactor Facility

    International Nuclear Information System (INIS)

    Park, Jaekwan; Suh, Yongsuk

    2014-01-01

    The primary purpose of the safety parameter monitoring system (SPDS) is to help operating personnel in the control room make quick assessments of the plant safety status. Thus, the basic function of the SPDS is a provision of a continuous indication of plant parameters or derived variables representative of the safety status of the plant. NUREG-0737 Supplement 1 provides details of the functional criteria for the SPDS, as one of the action plan requirements from TMI accident. The system provides various functions as follows: · Alerting based on safety function decision logics, · Success path analysis to achieve the integrity of the safety functions, · 3 layer display architecture - safety function, success path display for each safety function, system summary and equipment details for each safety function, · Integration with computer-based procedure. According to a Notice of the NSSC No. 2012-31, a research reactor facility generating more than 2 MW of power should also be furnished with the SPDS for emergency preparedness. Generally, a research reactor is a small size facility, and its number of instrumentations is fewer than that of NPPs. In particular, it is actually hard to have various and powerful functions from an economic perspective. Therefore, a safety parameter display system optimized for a research reactor facility must be proposed. This paper provides the requirement analysis results and proposes the design of safety parameter monitoring function for a research reactor. The safety parameter monitoring function supporting control room personnel during emergency conditions should be designed in a research reactor facility. The facility size and number of signals are smaller than that of the power plants. Also, it is actually hard to have various and powerful functions of nuclear power plants from an economic perspective. Thus, a safety parameter display system optimized to a research reactor must be proposed. First, we found important design items

  11. Design of Safety Parameter Monitoring Function in a Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jaekwan; Suh, Yongsuk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The primary purpose of the safety parameter monitoring system (SPDS) is to help operating personnel in the control room make quick assessments of the plant safety status. Thus, the basic function of the SPDS is a provision of a continuous indication of plant parameters or derived variables representative of the safety status of the plant. NUREG-0737 Supplement 1 provides details of the functional criteria for the SPDS, as one of the action plan requirements from TMI accident. The system provides various functions as follows: · Alerting based on safety function decision logics, · Success path analysis to achieve the integrity of the safety functions, · 3 layer display architecture - safety function, success path display for each safety function, system summary and equipment details for each safety function, · Integration with computer-based procedure. According to a Notice of the NSSC No. 2012-31, a research reactor facility generating more than 2 MW of power should also be furnished with the SPDS for emergency preparedness. Generally, a research reactor is a small size facility, and its number of instrumentations is fewer than that of NPPs. In particular, it is actually hard to have various and powerful functions from an economic perspective. Therefore, a safety parameter display system optimized for a research reactor facility must be proposed. This paper provides the requirement analysis results and proposes the design of safety parameter monitoring function for a research reactor. The safety parameter monitoring function supporting control room personnel during emergency conditions should be designed in a research reactor facility. The facility size and number of signals are smaller than that of the power plants. Also, it is actually hard to have various and powerful functions of nuclear power plants from an economic perspective. Thus, a safety parameter display system optimized to a research reactor must be proposed. First, we found important design items

  12. Thermal hydraulic tests for reactor safety system -Research on the improvement of nuclear safety-

    International Nuclear Information System (INIS)

    Chung, Moon Ki; Park, Chun Kyeong; Yang, Seon Kyu; Chung, Chang Hwan; Chun, Shee Yeong; Song, Cheol Hwa; Chun, Hyeong Gil; Chang, Seok Kyu; Chung, Heung Joon; Won, Soon Yeon; Cho, Yeong Ro; Kim, Bok Deuk; Min, Kyeong Ho

    1994-07-01

    The present research aims at the development of the thermal hydraulic verification test technology for the reactor safety system of the conventional and advanced nuclear power plant and the development of the advanced thermal hydraulic measuring techniques. (Author)

  13. FFTF [Fast Flux Test Facility] reactor shutdown system reliability reevaluation

    International Nuclear Information System (INIS)

    Pierce, B.F.

    1986-07-01

    The reliability analysis of the Fast Flux Test Facility reactor shutdown system was reevaluated. Failure information based on five years of plant operating experience was used to verify original reliability numbers or to establish new ones. Also, system modifications made subsequent to performance of the original analysis were incorporated into the reevaluation. Reliability calculations and sensitivity analyses were performed using a commercially available spreadsheet on a personal computer. The spreadsheet was configured so that future failures could be tracked and compared with expected failures. A number of recommendations resulted from the reevaluation including both increased and decreased surveillance intervals. All recommendations were based on meeting or exceeding existing reliability goals. Considerable cost savings will be incurred upon implementation of the recommendations

  14. Reactor cover gas monitoring at the Fast Flux Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Bechtold, R A; Holt, F E; Meadows, G E; Schenter, R E [Westinghouse Hanford Company, Richland, WA (United States)

    1987-07-01

    The Fast Flux Test Facility (FFTF) is a 400 megawatt (thermal) sodium cooled reactor designed for irradiation testing of fuels, materials and components for LMRs. It is operated by the Westinghouse Hanford Company for the U. S. Department of Energy on the government-owned Hanford reservation near Richland, Washington. The first 100 day operating cycle began in April 1982 and the eighth operating cycle was completed In July 1986. Argon is used as the cover gas for all sodium systems at the plant. A program for cover gas monitoring has been in effect since the start of sodium fill in 1978. The argon is supplied to the FFTF by a liquid argon Dewar System and used without further purification. A liquid argon Dewar system provides the large volume of inert gas required for operation of the FFTF. The gas is used as received and is not recycled. Low concentrations of krypton and xenon in the argon supply are essential to preclude interference with the gas tag system. Gas chromatography has been valuable for detection of inadvertent air in leakage during refueling operations. A temporary system is installed over the reactor during outages to prevent oxide formation in the sodium vapor traps upstream from the on line gas chromatograph. On line gas monitoring by gamma spectrometry and grab sampling with GTSTs has been successful for the identification of numerous radioactive gas releases from creep capsule experiments as well as 9 fuel pin ruptures. A redundant fission gas monitoring system has been installed to insure constant surveillance of the reactor cover gas.

  15. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-04-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: • Identifies pre-conceptual design requirements • Develops test loop equipment schematics and layout • Identifies space allocations for each of the facility functions, as required • Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems • Identifies pre-conceptual utility and support system needs • Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

  16. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    International Nuclear Information System (INIS)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-01-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: (1) Identifies pre-conceptual design requirements; (2) Develops test loop equipment schematics and layout; (3) Identifies space allocations for each of the facility functions, as required; (4) Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems; (5) Identifies pre-conceptual utility and support system needs; and (6) Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs

  17. The DRAGON aerosol research facility to study aerosol behaviour for reactor safety applications

    International Nuclear Information System (INIS)

    Suckow, Detlef; Guentay, Salih

    2008-01-01

    During a severe accident in a nuclear power plant fission products are expected to be released in form of aerosol particles and droplets. To study the behaviour of safety relevant reactor components under aerosol loads and prototypical severe accident conditions the multi-purpose aerosol generation facility DRAGON is used since 1994 for several projects. DRAGON can generate aerosol particles by the evaporation-condensation technique using a plasma torch system, fluidized bed and atomization of particles suspended in a liquid. Soluble, hygroscopic aerosol (i.e. CsOH) and insoluble aerosol particles (i.e. SnO 2 , TiO 2 ) or mixtures of them can be used. DRAGON uses state-of-the-art thermal-hydraulic, data acquisition and aerosol measurement techniques and is mainly composed of a mixing chamber, the plasma torch system, a steam generator, nitrogen gas and compressed air delivery systems, several aerosol delivery piping, gas heaters and several auxiliary systems to provide vacuum, coolant and off-gas treatment. The facility can be operated at system pressure of 5 bars, temperatures of 300 deg. C, flow rates of non-condensable gas of 900 kg/h and steam of 270 kg/h, respectively. A test section under investigation is attached to DRAGON. The paper summarizes and demonstrates with the help of two project examples the capabilities of DRAGON for reactor safety studies. (authors)

  18. Reactor safety research program. A description of current and planned reactor safety research sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research

    International Nuclear Information System (INIS)

    1975-06-01

    The reactor safety research program, sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, is described in terms of its program objectives, current status, and future plans. Elements of safety research work applicable to water reactors, fast reactors, and gas cooled reactors are presented together with brief descriptions of current and planned test facilities. (U.S.)

  19. Seismic qualification of safety class components in non-reactor nuclear facilities at Hanford site

    International Nuclear Information System (INIS)

    Ocoma, E.C.

    1989-01-01

    This paper presents the methods used during the walkdowns to compile as-built structural information to seismically qualify or verify the seismic adequacy of safety class components in the Plutonium Finishing Plant complex. The Plutonium finishing Plant is a non-reactor nuclear facility built during the 1950's and was designed to the Uniform Building Code criteria for both seismic and wind events. This facility is located at the US Department of Energy Hanford Site near Richland, Washington

  20. CSNI collective statement on support facilities for existing and advanced reactors. The function of OECD/Nea joint projects Nea committee on the safety of nuclear installations (CSNI)

    International Nuclear Information System (INIS)

    2008-01-01

    The NEA Committee on the Safety of Nuclear Installations (CSNI) has recently completed a study on the availability and utilisation of facilities supporting safety studies for current and advanced nuclear power reactors. The study showed that significant steps had been undertaken in the past several years in support of safety test facilities, mainly by conducting multinational joint projects centered on the capability of unique test facilities worldwide. Given the positive experience of the safety research projects, it has been recommended that efforts be made to prioritize technical issues associated with advanced (Generation IV) reactor designs and to develop options on how to efficiently obtain the necessary data through internationally co-ordinated research, preparing a gradual extension of safety research beyond the needs set by currently operating reactors. This statement constitutes a reference for future CSNI activities and for safety authorities, R and D centres and industry for internationally co-ordinated research initiatives in the nuclear safety research area. (author)

  1. Guide to the safety design examination about light water reactor facilities for power generation

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    This guide was compiled to evaluate the validity of the design policy when the safety design is examined at the time of the application for approval of the installation of nuclear reactors. About 7 years has elapsed since the existing guide was established, and the more appropriate guide to evaluate the safety should be made on the basis of the knowledge and experience accumulated thereafter. The range of application of this guide is limited to the above described evaluation, and it is not intended as the general standard for the design of nuclear reactors. First, the definition of the words used in this guide is given. Then, the guide to the safety examination is described about the general matters of reactor facilities, nuclear reactors and the measuring and controlling system, reactor-stopping system, reactivity-controlling system and safety protection system, reactor-cooling system, reactor containment vessels, fuel handling and waste treatment system. Several matters which require attention in the application of this guide or the clarification of the significance and interpretation of the guide itself were found, therefore the explanation about them was added at the end of this guide. (Kako, I.)

  2. Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)

    Energy Technology Data Exchange (ETDEWEB)

    Bradley K. Heath

    2014-03-01

    This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show that fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.

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

    International Nuclear Information System (INIS)

    G. L. Sharp; R. T. McCracken

    2004-01-01

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

  4. Safety test facilities. Needs and concepts. A French evaluation

    International Nuclear Information System (INIS)

    Tretiakoff, O.; Bailly, J.

    1976-01-01

    The fuel behaviour of LMFBRs in the event of an accident has been tested in-pile in the SCARABEE program (local blockage, sudden flow reduction and pump coast-down at constant power). These tests will be carried on in the framework of an international cooperation on irradiated fuels: this is the purpose of the CABRI and SCARABEE N programs. All those studies should enable to assess safety margins between accident conditions and the technical specifications of the reactor. The paper explains how a logical set of simple observations has led to the present state of the Cadarache in-pile experimental safety program and how it may help to find the way in a dense forest of both technical and psychological difficulties

  5. Safety test facilities. Needs and concepts. A French evaluation

    International Nuclear Information System (INIS)

    Tretiakoff, O.; Bailly, J.

    1976-01-01

    The fuel behavior of LMFBRs in the event of an accident has been tested in-pile in the SCARABEE program. These tests will be carried on in the framework of an international cooperation on irradiated fuels: this is the purpose of the CABRI and SCARABEE N programs. All those studies should enable to assess safety margins between accident conditions and the technical specifications of the reactor. The purpose of this paper is to explain how a logical set of simple observations has led us to the present state of the Cadarache in-pile experimental safety program and how it may help us to find our way in a dense forest of both technical and psychological difficulties

  6. Closed Loop In-Reactor Assembly (CLIRA): a fast flux test facility test vehicle

    International Nuclear Information System (INIS)

    Oakley, D.J.

    1978-01-01

    The Closed Loop In-Reactor Assembly (CLIRA) is a test vehicle for in-core material and fuel experiments in the Fast Flux Test Facility (FFTF). The FFTF is a fast flux nuclear test reactor operated for the Department of Energy (DOE) by Westinghouse Hanford Company in Richland, Washington. The CLIRA is a removable/replaceable part of the Closed Loop System (CLS) which is a sodium coolant system providing flow and temperature control independent of the reactor coolant system. The primary purpose of the CLIRA is to provide a test vehicle which will permit testing of nuclear fuels and materials at conditions more severe than exist in the FTR core, and to isolate these materials from the reactor core

  7. WWER type reactor primary loop imitation on large test loop facility in MARIA reactor

    International Nuclear Information System (INIS)

    Moldysh, A.; Strupchevski, A.; Kmetek, Eh.; Spasskov, V.P.; Shumskij, A.M.

    1982-01-01

    At present in Poland in cooperation with USSR a nuclear water loop test facility (WL) in 'MARIA' reactor in Sverke is under construction. The program objective is to investigate processes occuring in WWER reactor under emergency conditions, first of all after the break of the mainprimary loop circulation pipe-line. WL with the power of about 600 kW consists of three major parts: 1) an active loop, imitating the undamaged loops of the WWER reactor; 2) a passive loop assignedfor modelling the broken loop of the WWER reactor; 3) the emergency core cooling system imitating the corresponding full-scale system. The fuel rod bundle consists of 18 1 m long rods. They were fabricated according to the standard WWER fuel technology. In the report some general principles of WWERbehaviour imitation under emergency conditions are given. They are based on the operation experience obtained from 'SEMISCALE' and 'LOFT' test facilities in the USA. A description of separate modelling factors and criteria effects on the development of 'LOCA'-type accident is presented (the break cross-section to the primary loop volume ratio, the pressure differential between inlet and outlet reactor chambers, the pressure drop rate in the loop, the coolant flow rate throuh the core etc.). As an example a comparison of calculated flow rate variations for the WWER-1000 reactor and the model during the loss-of-coolant accident with the main pipe-line break at the core inlet is given. Calculations have been carried out with the use of TECH'-M code [ru

  8. Safety Research Experiment Facility Project. Conceptual design report. Volume V. Reactor vessel and closure

    International Nuclear Information System (INIS)

    1975-12-01

    The Prestressed Concrete Reactor Vessel (PCRV) will serve as the primary pressure retaining structure for the Safety Research Experiment Facility (SAREF) reactor. The reactor core, control rod drive room, primary heat exchangers, and gas circulators will be located in cavities within the PCRV. The orientation of these cavities, except for the control rod drive room, will be similar to the high-temperature gas-cooled reactor (HTGR) designs that are currently proposed or under design. Due to the nature of this type of structure, all biological and radiological shielding requirements are incorporated into the basic vessel design. At the midcore plane there are three radially oriented slots that will extend from the outside surface of the PCRV to the reactor core liner. These slots will accommodate each of the fuel motion monitoring systems which will be part of the observation apparatus used with the loop experiments

  9. Reactor facility

    International Nuclear Information System (INIS)

    Suzuki, Hiroaki; Murase, Michio; Yokomizo, Osamu.

    1997-01-01

    The present invention provides a BWR type reactor facility capable of suppressing the amount of steams generated by the mutual effect of a failed reactor core and coolants upon occurrence of an imaginal accident, and not requiring spacial countermeasures for enhancing the pressure resistance of the container vessel. Namely, a means for supplying cooling water at a temperature not lower by 30degC than the saturated temperature corresponding to the inner pressure of the containing vessel upon occurrence of an accident is disposed to a lower dry well below the pressure vessel. As a result, upon occurrence of such an accident that the reactor core should be melted and flown downward of the pressure vessel, when cooling water at a temperature not lower than the saturated temperature, for example, cooling water at 100degC or higher is supplied to the lower dry well, abrupt generation of steams by the mutual effect of the failed reactor core and cooling water is scarcely caused compared with a case of supplying cooling water at a temperature lower than the saturation temperature by 30degC or more. Accordingly, the amount of steams to be generated can be suppressed, and special countermeasure is no more necessary for enhancing the pressure resistance of the container vessel is no more necessary. (I.S.)

  10. Preliminary safety evaluation (PSE) for Sodium Storage Facility at the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Bowman, B.R.

    1994-01-01

    This evaluation was performed for the Sodium Storage Facility (SSF) which will be constructed at the Fast Flux Test Facility (FFTF) in the area adjacent to the South and West Dump Heat Exchanger (DHX) pits. The purpose of the facility is to allow unloading the sodium from the FFTF plant tanks and piping. The significant conclusion of this Preliminary Safety Evaluation (PSE) is that the only Safety Class 2 components are the four sodium storage tanks and their foundations. The building, because of its imminent risk to the tanks under an earthquake or high winds, will be Safety Class 3/2, which means the building has a Safety Class 3 function with the Safety Class 2 loads of seismic and wind factored into the design

  11. Development of a Test Facility to Simulate the Reactor Flow Distribution of APR+

    International Nuclear Information System (INIS)

    Euh, D. J.; Cho, S.; Youn, Y. J.; Kim, J. T.; Kang, H. S.; Kwon, T. S.

    2011-01-01

    Recently a design of new reactor, APR+, is being developed, as an advanced type of APR1400. In order to analyze the thermal margin and hydraulic characteristics of APR+, quantification tests for flow and pressure distribution with a conservation of flow geometry are necessary. Hetsroni (1967) proposed four principal parameters for a hydraulic model representing a nuclear reactor prototype: geometry, relative roughness, Reynolds number, and Euler number. He concluded that the Euler number should be similar in the prototype and model under the preservation of the aspect ratio on the flow path. The effect of the Reynolds number at its higher values on the Euler number is rather small, since the dependency of the form and frictional loss coefficients on the Reynolds number is seen to be small. ABB-CE has carried out several reactor flow model test programs, mostly for its prototype reactors. A series of tests were conducted using a 3/16 scale reactor model. (see Lee et al., 2001). Lee et al (1991) performed experimental studies using a 1/5.03 scale reactor flow model of Yonggwang nuclear units 3 and 4. They showed that the measured data met the acceptance criteria and were suitable for their intended use in terms of performance and safety analyses. The design of current test facility was based on the conservation of Euler number which is a ratio of pressure drop to dynamic pressure with a sufficiently turbulent region having a high Reynolds number. By referring to the previous study, the APR+ design is linearly reduced to 1/5 ratio with a 1/2 of the velocity scale, which yields a 1/39.7 of Reynolds number scaling ratio. In the present study, the design feature of the facilities, named 'ACOP', in order to investigate flow and pressure distribution are described

  12. TREAT [Transient Reactor Test Facility] reactor control rod scram system simulations and testing

    International Nuclear Information System (INIS)

    Solbrig, C.W.; Stevens, W.W.

    1990-01-01

    Air cylinders moving heavy components (100 to 300 lbs) at high speeds (above 300 in/sec) present a formidable end-cushion-shock problem. With no speed control, the moving components can reach over 600 in/sec if the air cylinder has a 5 ft stroke. This paper presents an overview of a successful upgrade modification to an existing reactor control rod drive design using a computer model to simulate the modified system performance for system design analysis. This design uses a high speed air cylinder to rapidly insert control rods (278 lb moved 5 ft in less than 300 msec) to scram an air-cooled test reactor. Included is information about the computer models developed to simulate high-speed air cylinder operation and a unique new speed control and end cushion design. A patent application is pending with the US Patent ampersand Trade Mark Office for this system (DOE case number S-68,622). The evolution of the design, from computer simulations thru operational testing in a test stand (simulating in-reactor operating conditions) to installation and use in the reactor, is also described. 6 figs

  13. The Fast Flux Test Facility built on safety

    International Nuclear Information System (INIS)

    1989-01-01

    No other high-tech industry has grown as fast as the nuclear industry. The information available to the general public has not kept pace with the rapid growth of nuclear data---its growth has outpaced its media image and the safety of nuclear facilities has become a highly debated issue. This book is an attempt to bridge the gap between the high-tech information of the nuclear industry and its understanding by the general public. It explains the three levels of defense at the Fast Flux Test Facility (FFTF) and why these levels provide an acceptable margin to protect the general public and on-site personnel, while achieving FFTF's mission to provide research and development for the US Department of Energy

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

    International Nuclear Information System (INIS)

    Duckwitz, Noel

    2011-01-01

    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.

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

    International Nuclear Information System (INIS)

    Duckwitz, Noel

    2011-01-01

    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.

  16. Ohmically heated toroidal experiment (OHTE) mobile ignition test reactor facility concept study

    International Nuclear Information System (INIS)

    Masson, L.S.; Watts, K.D.; Piscitella, R.R.; Sekot, J.P.; Drexler, R.L.

    1983-02-01

    This report presents the results of a study to evaluate the use of an existing nuclear test complex at the Idaho National Engineering Laboratory (INEL) for the assembly, testing, and remote maintenance of the ohmically heated toroidal experiment (OHTE) compact reactor. The portable reactor concept is described and its application to OHTE testing and maintenance requirements is developed. Pertinent INEL facilities are described and several test system configurations that apply to these facilities are developed and evaluated

  17. Research on the improvement of nuclear safety -Thermal hydraulic tests for reactor safety system-

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Moon Kee; Park, Choon Kyung; Yang, Sun Kyoo; Chun, Se Yung; Song, Chul Hwa; Jun, Hyung Kil; Jung, Heung Joon; Won, Soon Yun; Cho, Yung Roh; Min, Kyung Hoh; Jung, Jang Hwan; Jang, Suk Kyoo; Kim, Bok Deuk; Kim, Wooi Kyung; Huh, Jin; Kim, Sook Kwan; Moon, Sang Kee; Lee, Sang Il [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-06-01

    The present research aims at the development of the thermal hydraulic verification test technology for the safety system of the conventional and advanced nuclear power plant and the development of the advanced thermal hydraulic measuring techniques. In this research, test facilities simulating the primary coolant system and safety system are being constructed for the design verification tests of the existing and advanced nuclear power plant. 97 figs, 14 tabs, 65 refs. (Author).

  18. Testing of the multi-application small light water reactor (MASLWR) passive safety systems

    International Nuclear Information System (INIS)

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

    2007-01-01

    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). The MASLWR is a pressurized light water reactor design with a net output of 35 MWe that uses natural circulation in both normal and transient operation. Due to its small size, portability and modularity, the MASLWR design is well suited to help fill the potential need for grid appropriate reactor designs for smaller electricity grids as may be found in developing or remote regions. The purpose of the OSU MASLWR test facility is to assess the operation of the MASLWR under normal full operating 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. During the performance of these tests, plant operations to include start up, normal operation and shut down evolutions were demonstrated successfully

  19. Very high temperature measurements: Applications to nuclear reactor safety tests

    International Nuclear Information System (INIS)

    Parga, Clemente-Jose

    2013-01-01

    This PhD dissertation focuses on the improvement of very high temperature thermometry (1100 deg. C to 2480 deg. C), 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 (±0.001 deg. C) to applied research with a reasonable degradation of uncertainties (±3-5 deg. C). - The corrosion study and metallurgical characterization of Type-C thermocouple (service temp. 2300 deg. C) prospective sheath material was undertaken to extend the survivability of TCs used for molten metallic/oxide corium thermometry (below 2000 deg. C)

  20. Safety analysis of the 700-horsepower combustion test facility

    Energy Technology Data Exchange (ETDEWEB)

    Berkey, B.D.

    1981-05-01

    The objective of the program reported herein was to provide a Safety Analysis of the 700 h.p. Combustion Test Facility located in Building 93 at the Pittsburgh Energy Technology Center. Extensive safety related measures have been incorporated into the design, construction, and operation of the Combustion Test Facility. These include: nitrogen addition to the coal storage bin, slurry hopper, roller mill and pulverizer baghouse, use of low oxygen content combustion gas for coal conveying, an oxygen analyzer for the combustion gas, insulation on hot surfaces, proper classification of electrical equipment, process monitoring instrumentation and a planned remote television monitoring system. Analysis of the system considering these factors has resulted in the determination of overall probabilities of occurrence of hazards as shown in Table I. Implementation of the recommendations in this report will reduce these probabilities as indicated. The identified hazards include coal dust ignition by hot ductwork and equipment, loss of inerting within the coal conveying system leading to a coal dust fire, and ignition of hydrocarbon vapors or spilled oil, or slurry. The possibility of self-heating of coal was investigated. Implementation of the recommendations in this report will reduce the ignition probability to no more than 1 x 10/sup -6/ per event. In addition to fire and explosion hazards, there are potential exposures to materials which have been identified as hazardous to personal health, such as carbon monoxide, coal dust, hydrocarbon vapors, and oxygen deficient atmosphere, but past monitoring experience has not revealed any problem areas. The major environmental hazard is an oil spill. The facility has a comprehensive spill control plan.

  1. Gas cooled fast breeder reactor design for a circulator test facility (modified HTGR circulator test facility)

    Energy Technology Data Exchange (ETDEWEB)

    1979-10-01

    A GCFR helium circulator test facility sized for full design conditions is proposed for meeting the above requirements. The circulator will be mounted in a large vessel containing high pressure helium which will permit testing at the same power, speed, pressure, temperature and flow conditions intended in the demonstration plant. The electric drive motor for the circulator will obtain its power from an electric supply and distribution system in which electric power will be taken from a local utility. The conceptual design decribed in this report is the result of close interaction between the General Atomic Company (GA), designer of the GCFR, and The Ralph M. Parson Company, architect/engineer for the test facility. A realistic estimate of total project cost is presented, together with a schedule for design, procurement, construction, and inspection.

  2. Reports and operational engineering: An independent safety assessment of Department of Energy nuclear reactor facilities

    International Nuclear Information System (INIS)

    Rochman, A.; Washburn, B.W.

    1981-02-01

    The Nuclear Facilities Personnel Qualification and Training (NFPQT) Committee, established via an October 24, 1979 memorandum from the Department of Energy (DOE) Under Secretary, was instructed to review the ''Kemeny Commission'' recommendations and to identify possible implications for DOE's nuclear facilities. As a result of this review, the Committee recommended that DOE carry out assessments in seven categories. The assessments would address specific topics identified for each category as delineated in the NFPQT ''Guidelines for Assessing the Safe Operation of DOE-Owned Reactors,'' dated May 7, 1980. The Committee recognized that similar assessments had been ongoing in the DOE program and safety overview organizations since the Three Mile Island nuclear accident and it was the Committee's intent to use the results of those ongoing assessments as an input to their evaluations. This information would be supplemented by additional studies consisting of the subject-related documents used at each reactor facility studied, and an on-site review of these reactor facilities by professional personnel within the Department of Energy, its operating contractors and independent consultants. 1 tab

  3. Russian Federation: Passive Safety Components for Lead-Cooled Reactor Facilities

    International Nuclear Information System (INIS)

    Sarkulov, M.K.

    2015-01-01

    There is a specific range of engineered features used traditionally in nuclear technology. As a rule, main reactivity control systems use conventional active actuators with solid-body control members and/or liquid systems with active injection of liquid absorber. Other operation principles are normally chosen for additional systems. Currently, the traditional approach to improving the reliability of a reactor facility suggests an increase in the number of safety components and systems which provide for mutual assurance or assist each other. There is a great variety of additional reactivity control members designed for the reactor facility control and shutdown, including hydrodynamic members in the form of rods (acting from the coolant flow); floating-type members (absorbers and displacers); storage-type and liquid members (used in separate channels); bulk members (pebble absorber); gas-based members (with a gas absorber); shape-memory members and others. Hydrodynamic systems were introduced at Beloyarsk NPP Units 1 and 2 and proposed for use in other facility designs, Gases and bulk materials have not been commonly accepted: the former because of the high cost of high-efficiency gaseous absorbers, and the latter because of the complecated monitoring of the bulk material position. It is rather difficult and not always necessary to use the same engineering approaches in new lead-cooled reactor facilities as in traditional ones. Similarly to the development of traditional safety systems, passive safety components (devices) shall be designed according to the essential requirements of the nuclear regulations of the Russian Federation

  4. Fast flux test facility final safety analysis report amendment 79

    International Nuclear Information System (INIS)

    Dautel, W.A.

    1999-01-01

    This document is provided to replace, remove, or add applicable pages to the chapters on: Heat Transport System; Containment and Structures; Auxiliary Systems; Reactor Refueling System; Conduct of Operations; Safety Analysis; Quality Assurance; FFTF Criticality Specifications; and Appendix H's TRIGA Fuel Storage System

  5. Acoustic detection of boiling in the Sodium Loop Safety Facility in-reactor experiment P1

    International Nuclear Information System (INIS)

    Carey, W.M.; Anderson, T.T.; Bobis, J.P.

    1976-06-01

    Acoustic data were obtained from two high-temperature lithium niobate microphones on the loop background noise and transient pressure pulses during the Sodium Loop Safety Facility (SLSF) P1 in-reactor experiment. This experiment simulated an LMFBR loss-of-piping-integrity (LOPI) transient on a nineteen element, end-of-life, enriched-UO 2 fuel assembly. The microphones were exposed to liquid sodium at a distance 4.85 meters above the reactor core at temperatures between 315 0 and 590 0 C. The distance and location of the microphones in the P1 Test Train provided an attenuative transmission path which was undesirable for optimum acoustic detection of sodium boiling and fuel failure. The data gathered on the loop background noise was observed to be dominated by pump and electrical noise at frequencies below 1.5 KHz and appeared to be dominated by flow induced local turbulence noise at higher frequencies. During the period of time that the sodium in the fuel assembly was at its saturation temperature 943 0 C (1730 0 F), as indicated by the wire wrap thermocouples, several discrete pulses were observed with peak-to-peak pressure between 3.3 kPa and 7.9 kPa and center frequencies between 360 and 550 Hz. The pulses occurred at two separate gradually increasing repetition rates. These observations appear to be consistent with the result of an impulsive forcing function interacting with a band passed Helmholtz resonator. These data are consistent with the hypothesis that sodium boiling occurred in the P1 fuel assembly, resulting in the formation of individual voids that collapsed upon reaching the subcooled sodium. These data provide pertinent information regarding the feasibility of sodium boiling detection and may provide additional insight into the dynamics of the void behavior

  6. Two important safety-related verification tests in the design of Qinshan NPP 600 MWe reactor

    International Nuclear Information System (INIS)

    Li Pengzhou; Li Tianyong; Yu Danping; Sun Lei

    2005-01-01

    This paper summarizes two most important verification tests performed in the design of reactor of Qinshan NPP Phase II: seismic qualification test of control rod drive line (CRDL), flow-induced vibration test of reactor internals both in 1:5 scaled model and on-site measurement during heat function testing (HFT). Both qualification tests proved that the structural design of the reactor has large safety margin. (authors)

  7. On exposure of workers in nuclear reactor facilities for test and in nuclear reactor facilities in research and development stage in fiscal 1988

    International Nuclear Information System (INIS)

    1989-01-01

    The Law for Regulation on Nuclear Reactor requires the operators of nuclear reactors that the exposure dose of workers engaged in work for nuclear reactors should not exceed the limits specified in official notices that are issued based on the Law. The present article summarizes the contents of the Report on Radiation Management in 1988 submitted by the operators of nuclear reactor facilities for test and those of nuclear reactor facilities in research and development stage based on the Law, and the Report on Management of Exposure Dose of Workers submitted by them based on administrative notices. The reports demonstrate that the exposure of workers was below the permissible exposure dose in 1988 in all nuclear reactor facilities. The article presents data on the distribution of exposure dose among workers in all facilities with a nuclear reactor for test, and data on personal exposure of employees and non-employees and overall exposure of all workers in the facilities of Japan Atomic Energy Research Institute and Power Reactor and Nuclear Fuel Development Corporation. (N.K.)

  8. Reactor safety

    International Nuclear Information System (INIS)

    Meneley, D.A.

    The people of Ontario have begun to receive the benefits of a low cost, assured supply of electrical energy from CANDU nuclear stations. This indigenous energy source also has excellent safety characteristics. Safety has been one of the central themes of the CANDU development program from its very beginning. A great deal of work has been done to establish that public risks are small. However, safety design criteria are now undergoing extensive review, with a real prospect of more stringent requirements being applied in the future. Considering the newness of the technology it is not surprising that a consensus does not yet exist; this makes it imperative to discuss the issues. It is time to examine the policies and practice of reactor safety management in Canada to decide whether or not further restrictions are justified in the light of current knowledge

  9. Safety analysis calculations for research and test reactors

    International Nuclear Information System (INIS)

    Chen, S.Y.; MacDonald, R.; MacFarlane, D.

    1983-01-01

    Safety issues for the two general types of reactors, i.e., the plate-type (MTR-type) reactor and the rod-type (TRIGA-type) reactor, resulting from the changes associated with LEU vs HEU fuels, are explored. The plate-type fuels are typically uranium aluminide (UAl/sub x/) compounds dispersed in aluminum and clad with aluminum. Moderation is provided by the water coolant. Self shut-down reactivity coefficients with HEU fuel are entirely a result of coolant heating, whereas with LEU fuel there is an additional shut down contribution provided by the direct heating of the fuel due to the Doppler coefficient. In contrast, the rod-type (TRIGA) fuels are mixtures of zirconium hydride, uranium, and erbium. This fuel mixture is formed into rods (approx. 1 cm diameter) and clad with stainless steel or Incoloy. In the TRIGA fuel the self-shutdown reactivity is more complex, depending on heating of the fuel rather than the coolant. Results of transient calculations performed with existing computer codes, most suited for each type of reactor, are presented

  10. Utilizing the Fast Flux Test Facility for international passive safety testing

    International Nuclear Information System (INIS)

    Shen, P.K.; Padilla, A.; Lucoff, D.M.; Waltar, A.E.

    1991-01-01

    A two-phased approach has been undertaken in the Fast Flux Test Facility (FFTF) to conduct passive safety testing. Phase I (1986 to 1987) was structured to obtain an initial understanding of the reactivity feedback components. The planned Phase II (1992 to 1993) international program will extend the testing to include static and dynamic feedback measurements, transient and demonstration tests, and gas expansion module (GEM) reactivity tests. The primary objective is to meet the needs for safety analysis code validation, with particular emphasis on reducing the uncertainties associated with structure reactivity feedback. Program scope and predicted FFTF responses are discussed and illustrated. (author)

  11. Conceptual design of a fission-based integrated test facility for fusion reactor components

    International Nuclear Information System (INIS)

    Watts, K.D.; Deis, G.A.; Hsu, P.Y.S.; Longhurst, G.R.; Masson, L.S.; Miller, L.G.

    1982-01-01

    The testing of fusion materials and components in fission reactors will become increasingly important because of lack of fusion engineering test devices in the immediate future and the increasing long-term demand for fusion testing when a fusion reactor test station becomes available. This paper presents the conceptual design of a fission-based Integrated Test Facility (ITF) developed by EG and G Idaho. This facility can accommodate entire first wall/blanket (FW/B) test modules such as those proposed for INTOR and can also accommodate smaller cylindrical modules similar to those designed by Oak Ridge National laboratory (ORNL) and Westinghouse. In addition, the facility can be used to test bulk breeder blanket materials, materials for tritium permeation, and components for performance in a nuclear environment. The ITF provides a cyclic neutron/gamma flux as well as the numerous module and experiment support functions required for truly integrated tests

  12. Advanced Test Reactor National Scientific User Facility 2010 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Mary Catherine Thelen; Todd R. Allen

    2011-05-01

    This is the 2010 ATR National Scientific User Facility Annual Report. This report provides an overview of the program for 2010, along with individual project reports from each of the university principal investigators. The report also describes the capabilities offered to university researchers here at INL and at the ATR NSUF partner facilities.

  13. On exposure management of workers in nuclear reactor facilities for test and in nuclear reactor facilities in research and development stage in fiscal 1993

    International Nuclear Information System (INIS)

    1994-01-01

    The Law of Regulation on Nuclear Reactor requires the operators of nuclear reactors that the exposure dose of workers engaged in work for nuclear reactors should not exceed the limits specified in official notices that are issued based on the Law. The present article summarizes the contents of the Report on Radiation Management in 1993 submitted by the operators of nuclear reactor facilities for test and those of nuclear reactor facilities in research and development stage based on the Law, and the Report on Management of Exposure Dose of Workers submitted by them based on administrative notices. The reports demonstrate that the the exposure of workers was below the permissible exposure dose in 1993 in all nuclear reactor facilities. The article presents data on the distribution of exposure dose among workers in all facilities with a nuclear reactor for test, and data on personal exposure of employees and non-employees and overall exposure of all workers in the facilities of JAERI and PNC. (J.P.N.)

  14. The Analysis of Loop Seal Purge Time for the KHNP Pressurizer Safety Valve Test Facility Using the GOTHIC Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Ae; Kim, Chang Hyun; Kweon, Gab Joo; Park, Jong Woon [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

    2007-10-15

    The pressurizer safety valves (PSV) in Pressurized Water Reactors are required to provide the overpressure protection for the Reactor Coolant System (RCS) during the overpressure transients. Korea Hydro and Nuclear Power Company (KHNP) plans to build the PSV test facility for the purpose of providing the PSV pop-up characteristics and the loop seal dynamics for the new safety analysis. When the pressurizer safety valve is mounted in a loop seal configuration, the valve must initially pass the loop seal water prior to popping open on steam. The loop seal in the upstream of PSV prevents leakage of hydrogen gas or steam through the safety valve seat. This paper studies on the loop seal clearing dynamics using GOTHIC-7.2a code to verify the effects of loop seal purge time on the reactor coolant system overpressure.

  15. Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Munn, W.I.

    1981-01-01

    The Fast Flux Test Facility (FFTF), located on the Hanford site a few miles north of Richland, Washington, is a major link in the chain of development required to sustain and advance Liquid Metal Fast Breeder Reactor (LMFBR) technology in the United States. This 400 MWt sodium cooled reactor is a three loop design, is operated by Westinghouse Hanford Company for the US Department of Energy, and is the largest research reactor of its kind in the world. The purpose of the facility is three-fold: (1) to provide a test bed for components, materials, and breeder reactor fuels which can significantly extend resource reserves; (2) to produce a complete body of base data for the use of liquid sodium in heat transfer systens; and (3) to demonstrate inherent safety characteristics of LMFBR designs

  16. First TREAT [Transient Reactor Test Facility] transient overpower tests on U-Pu-Zr fuel: M5 and M6

    International Nuclear Information System (INIS)

    Robinson, W.R.; Bauer, T.H.; Wright, A.E.; Rhodes, E.A.; Stanford, G.S.; Klickman, A.E.

    1987-01-01

    Transient Reactor Test Facility (TREAT) tests M5 and M6 were the first transient overpower (TOP) tests of the margin to cladding breach and prefailure elongation of metallic U-Pu-Zr ternary fuel, the reference fuel of the Integral Fast Reactor concept. Similar tests on U-Fs fueled EBR-II driver pins were previously performed and reported [1,2]. Results from these earlier tests indicated a margin to failure of about 4 times nominal power and significant axial elongation prior to failure, a feature that was very pronounced at low burnups. While these two fuel types are similar in many respects, the ternary alloy exhibits a much more complex physical structure and is typically irradiated at much higher temperatures. Thus, a prime motivation for performing M5 and M6 was to compare the safety related fuel performance characteristics of U-Fs and U-Pu-Zr. This report described conditions, results, and conclusions of testing of these fuel types

  17. Concerning control of radiation exposure to workers in nuclear reactor facilities for testing and nuclear reactor facilities in research and development phase (fiscal 1987)

    International Nuclear Information System (INIS)

    1988-01-01

    A nuclear reactor operator is required by the Nuclear Reactor Control Law to ensure that the radiation dose to workers engaged in the operations of his nuclear reactor is controlled below the permissible exposure doses that are specified in notifications issued based on the Law. The present note briefly summarizes the data given in the Reports on Radiation Control, which have been submitted according to the Nuclear Reactor Control Law by the operators of nuclear reactor facilities for testing and those in the research and development phase, and the Reports on Control of Radiation Exposure to Workers submitted in accordance with the applicable administrative notices. According to these reports, the measured exposure to workers in 1987 were below the above-mentioned permissible exposure doses in all these nuclear facilities. The 1986 and 1987 measurements of radiation exposure dose to workers in nuclear reactor facilities for testing are tabulated. The measurements cover dose distribution among the facilities' personnel and workers of contractors. They also cover the total exposure dose for all workers in each of four plants operated under the Japan Atomic Energy Research Institute and the Power Reactor and Nuclear Fuel Development Corporation. (N.K.)

  18. The approaches of safety design and safety evaluation at HTTR (High Temperature Engineering Test Reactor)

    International Nuclear Information System (INIS)

    Iigaki, Kazuhiko; Saikusa, Akio; Sawahata, Hiroaki; Shinozaki, Masayuki; Tochio, Daisuke; Honma, Fumitaka; Tachibana, Yukio; Iyoku, Tatsuo; Kawasaki, Kozo; Baba, Osamu

    2006-06-01

    Gas Cooled Reactor has long history of nuclear development, and High Temperature Gas Cooled Reactor (HTGR) has been expected that it can be supply high temperature energy to chemical industry and to power generation from the points of view of the safety, the efficiency, the environment and the economy. The HTGR design is tried to installed passive safety equipment. The current licensing review guideline was made for a Low Water Reactor (LWR) on safety evaluation therefore if it would be directly utilized in the HTGR it needs the special consideration for the HTGR. This paper describes that investigation result of the safety design and the safety evaluation traditions for the HTGR, comparison the safety design and safety evaluation feature for the HTGT with it's the LWR, and reflection for next HTGR based on HTTR operational experiment. (author)

  19. Accident safety analysis for 300 Area N Reactor Fuel Fabrication and Storage Facility

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, D.J.; Brehm, J.R.

    1994-01-01

    The purpose of the accident safety analysis is to identify and analyze a range of credible events, their cause and consequences, and to provide technical justification for the conclusion that uranium billets, fuel assemblies, uranium scrap, and chips and fines drums can be safely stored in the 300 Area N Reactor Fuel Fabrication and Storage Facility, the contaminated equipment, High-Efficiency Air Particulate filters, ductwork, stacks, sewers and sumps can be cleaned (decontaminated) and/or removed, the new concretion process in the 304 Building will be able to operate, without undue risk to the public, employees, or the environment, and limited fuel handling and packaging associated with removal of stored uranium is acceptable.

  20. Accident safety analysis for 300 Area N Reactor Fuel Fabrication and Storage Facility

    International Nuclear Information System (INIS)

    Johnson, D.J.; Brehm, J.R.

    1994-01-01

    The purpose of the accident safety analysis is to identify and analyze a range of credible events, their cause and consequences, and to provide technical justification for the conclusion that uranium billets, fuel assemblies, uranium scrap, and chips and fines drums can be safely stored in the 300 Area N Reactor Fuel Fabrication and Storage Facility, the contaminated equipment, High-Efficiency Air Particulate filters, ductwork, stacks, sewers and sumps can be cleaned (decontaminated) and/or removed, the new concretion process in the 304 Building will be able to operate, without undue risk to the public, employees, or the environment, and limited fuel handling and packaging associated with removal of stored uranium is acceptable

  1. Integrated leak rate testing of the fast flux test facility reactor containment building

    International Nuclear Information System (INIS)

    James, E.B.; Farabee, O.A.; Bliss, R.J.

    1978-01-01

    The initial Integrated Leak Rate Test (ILRT) of the Fast Flux Test Facility containment building was performed from May 27 to June 2, 1978. The test was conducted in air with systems vented and with the containment recirculating coolers in operation. 10 psig and 5 psig tests were run using the absolute pressure test method. The measured leakage rates were .033% Vol/24 hr. and -.0015% Vol/24 hrs. respectively. Subsequent verification tests at both 10 psig and 5 psig proved that the test equipment was operating properly and it was sensitive enough to detect leaks at low pressures. This ILRT was performed at a lower pressure than any previous ILRT on a reactor containment structure in the United States. While the initial design requirements for ice condenser containments called for a part pressure test at 6 psig, the tests were waived due to the apparent statistical problems of data analysis and the repeatability of the data itself at such low pressure. In contrast to this belief, both the 5 and 10 psig ILRT's were performed in a successful manner at FFTF

  2. An independent safety assessment of Department of Energy nuclear reactor facilities: Procedures, operations and maintenance

    International Nuclear Information System (INIS)

    Toto, G.; Lindgren, A.J.

    1981-02-01

    The 1979 accident at the Three Mile Island commercial nuclear power plant has led to a number of studies of nuclear reactors, in both the public and private sectors. One of these is that of the Department of Energy's (DOE) Nuclear Facilities Personnel Qualification and Training (NFPQT) Committee, which has outlined tasks for assessment of 13 reactors owned by DOE and operated by contractors. This report covers one of the tasks, the assessment of procedures, operations, and maintenance at the DOE reactor facilities, based on a review of actual documents used at the reactor sites

  3. Fast Flux Test Facility: a unique resource for breeder reactor development

    International Nuclear Information System (INIS)

    Peckinpaugh, C.L.; Patterson, J.R.

    1984-08-01

    Based on encouraging data from the first two years of operation, the FFTF program has been redirected to close out testing on the intermediate lifetime fuel system and move directly to an early demonstration of a fuel system with a projected life of three to five years. Full advantage of favorable reactor physics effects will be taken to enhance inherent shutdown features and maximize reactor safety while reducing capital costs. Plant reliability will be demonstrated by achievement of high plant capacity factors

  4. Catalogue and classification of technical safety standards, rules and regulations for nuclear power reactors and nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    Fichtner, N.; Becker, K.; Bashir, M.

    1977-01-01

    The present report is an up-dated version of the report 'Catalogue and Classification of Technical Safety Rules for Light-water Reactors and Reprocessing Plants' edited under code No EUR 5362e, August 1975. Like the first version of the report, it constitutes a catalogue and classification of standards, rules and regulations on land-based nuclear power reactors and fuel cycle facilities. The reasons for the classification system used are given and discussed

  5. Technical safety requirements for the Annular Core Research Reactor Facility (ACRRF)

    International Nuclear Information System (INIS)

    Boldt, K.R.; Morris, F.M.; Talley, D.G.; McCrory, F.M.

    1998-01-01

    The Technical Safety Requirements (TSR) document is prepared and issued in compliance with DOE Order 5480.22, Technical Safety Requirements. The bases for the TSR are established in the ACRRF Safety Analysis Report issued in compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports. The TSR identifies the operational conditions, boundaries, and administrative controls for the safe operation of the facility

  6. Abbreviated sampling and analysis plan for planning decontamination and decommissioning at Test Reactor Area (TRA) facilities

    International Nuclear Information System (INIS)

    1994-10-01

    The objective is to sample and analyze for the presence of gamma emitting isotopes and hazardous constituents within certain areas of the Test Reactor Area (TRA), prior to D and D activities. The TRA is composed of three major reactor facilities and three smaller reactors built in support of programs studying the performance of reactor materials and components under high neutron flux conditions. The Materials Testing Reactor (MTR) and Engineering Test Reactor (ETR) facilities are currently pending D/D. Work consists of pre-D and D sampling of designated TRA (primarily ETR) process areas. This report addresses only a limited subset of the samples which will eventually be required to characterize MTR and ETR and plan their D and D. Sampling which is addressed in this document is intended to support planned D and D work which is funded at the present time. Biased samples, based on process knowledge and plant configuration, are to be performed. The multiple process areas which may be potentially sampled will be initially characterized by obtaining data for upstream source areas which, based on facility configuration, would affect downstream and as yet unsampled, process areas. Sampling and analysis will be conducted to determine the level of gamma emitting isotopes and hazardous constituents present in designated areas within buildings TRA-612, 642, 643, 644, 645, 647, 648, 663; and in the soils surrounding Facility TRA-611. These data will be used to plan the D and D and help determine disposition of material by D and D personnel. Both MTR and ETR facilities will eventually be decommissioned by total dismantlement so that the area can be restored to its original condition

  7. A simulated test of physical starting and reactor physics on zero power facility of PWR

    International Nuclear Information System (INIS)

    Yao Zewu; Ji Huaxiang; Chen Zhicheng; Yao Zhiquan; Chen Chen; Li Yuwen

    1995-01-01

    The core neutron economics has been verified through experiments conducted at a zero power reactor with baffles of various thickness. A simulated test of physical starting of Qinshan PWR has been introduced. The feasibility and safety of the programme are verified. The research provides a valuable foundation for developing physical starting programme

  8. Advancing nuclear technology and research. The advanced test reactor national scientific user facility

    Energy Technology Data Exchange (ETDEWEB)

    Benson, Jeff B; Marshall, Frances M [Idaho National Laboratory, Idaho Falls, ID (United States); Allen, Todd R [Univ. of Wisconsin, Madison, WI (United States)

    2012-03-15

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research. The mission of the ATR NSUF is to provide access to world-class facilities, thereby facilitating the advancement of nuclear science and technology. Cost free access to the ATR, INL post irradiation examination facilities, and partner facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to United States Department of Energy. To increase overall research capability, ATR NSUF seeks to form strategic partnerships with university facilities that add significant nuclear research capability to the ATR NSUF and are accessible to all ATR NSUF users. (author)

  9. GERDA test facility for pressurized water reactors with straight tube steam generators

    International Nuclear Information System (INIS)

    Ahrens, G.; Haury, G.; Lahner, K.; Schatz, A.

    1983-01-01

    A number of large-scale experimental facilities have been constructed and operate in order to experiment on the thermodynamic and thermohydraulic behaviour of nuclear facilities in case of LOCA. Most of them were designed for ''large leak'' accidents, but as ''small leak'' accidents became the focus of interest, such experiments were also carried out. Experiments carried out with this arrangement for PWR-type reactors with straight-tube steam generators are only partially evaluable. BBR and B and W therefore cooperated in the construction of the test facility GERDA, designed for testing reactors of BBR design. It supplied relevant experimental results for the nuclear power plant at Muelheim-Kaerlich. (orig.) [de

  10. Safety inspections to TRIGA reactors

    International Nuclear Information System (INIS)

    Byszewski, W.

    1988-01-01

    The operational safety advisory programme was created to provide useful assistance and advice from an international perspective to research reactor operators and regulators on how to enhance operational safety and radiation protection on their reactors. Safety missions cover not only the operational safety of reactors themselves, but also the safety of associated experimental loops, isotope laboratories and other experimental facilities. Safety missions are also performed on request in other Member States which are interested in receiving impartial advice and assistance in order to enhance the safety of research reactors. The results of the inspections have shown that in some countries there are problems with radiation protection practices and nuclear safety. Very often the Safety Analysis Report is not updated, regulatory supervision needs clarification and improvement, maintenance procedures should be more formalised and records and reports are not maintained properly. In many cases population density around the facility has increased affecting the validity of the original safety analysis

  11. Public's right to information: An independent safety assessment of Department of Energy nuclear reactor facilities

    International Nuclear Information System (INIS)

    Stokely, E.

    1981-02-01

    The events at TMI prompted the Under Secretary of the Department of Energy (DOE) to establish the Nuclear Facilities Personnel Qualification and Training (NFPQT) Committee. This Committee was assigned the task of assessing the adequacy of nuclear facility personnel qualification and training at DOE-owned reactors in light of the Three Mile Island accident. The Committee was also asked to review recommendations and identify possible implications for DOE's nuclear facilities

  12. The operation of the Tokamak Fusion Test Reactor Tritium Facility

    International Nuclear Information System (INIS)

    Gentile, C.A.; LaMarche, P.H.

    1995-01-01

    The TFTR tritium operations staff has successfully received, stored, handled, and processed over five hundred thousand curies of tritium for the purpose of supporting D-T (Deuterium-Tritium) operations at TFTR. Tritium operations personnel nominally provide continuous round the clock coverage (24 hours/day, 7 days/week) in shift complements consisting of I supervisor and 3 operators. Tritium Shift Supervisors and operators are required to have 5 years of operational experience in either the nuclear or chemical industry and to become certified for their positions. The certification program provides formal instruction, as well as on the job training. The certification process requires 4 to 6 months to complete, which includes an oral board lasting up to 4 hours at which time the candidate is tested on their knowledge of Tritium Technology and TFTR Tritium systems. Once an operator is certified, the training process continues with scheduled training weeks occurring once every 5 weeks. During D-T operations at TFTR the operators must evacuate the tritium area due to direct radiation from TFTR D-T pulses. During '' time operators maintain cognizance over tritium systems via a real time TV camera system. Operators are able to gain access to the Tritium area between TFTR D-T pulses, but have been excluded from die tritium area during D-T pulsing for periods up to 30 minutes. Tritium operators are responsible for delivering tritium gas to TFRR as well as processing plasma exhaust gases which lead to the deposition of tritium oxide on disposable molecular sieve beds (DMSB). Once a DMSB is loaded, the operations staff remove the expended DMSB, and replace it with a new DMSB container. The TFIR tritium system is operated via detailed procedures which require operator sign off for system manipulation. There are >300 procedures controlling the operation of the tritium systems

  13. Non Nuclear Testing of Reactor Systems In The Early Flight Fission Test Facilities (EFF-TF)

    International Nuclear Information System (INIS)

    Van Dyke, Melissa; Martin, James

    2004-01-01

    The Early Flight Fission-Test Facility (EFF-TF) can assist in the design and development of systems through highly effective non-nuclear testing of nuclear systems when technical issues associated with near-term space fission systems are 'non-nuclear' in nature (e.g. system's nuclear operations are understood). For many systems, thermal simulators can be used to closely mimic fission heat deposition. Axial power profile, radial power profile, and fuel pin thermal conductivity can be matched. In addition to component and subsystem testing, operational and lifetime issues associated with the steady state and transient performance of the integrated reactor module can be investigated. Instrumentation at the EFF-TF allows accurate measurement of temperature, pressure, strain, and bulk core deformation (useful for accurately simulating nuclear behavior). Ongoing research at the EFF-TF is geared towards facilitating research, development, system integration, and system utilization via cooperative efforts with DOE laboratories, industry, universities, and other Nasa centers. This paper describes the current efforts for the latter portion of 2003 and beginning of 2004. (authors)

  14. Experimental facilities for PEC reactor design central channel test loop: CPC-1 - thermal shocks loop: CEDI

    International Nuclear Information System (INIS)

    Calvaresi, C.; Moreschi, L.F.

    1983-01-01

    PEC (Prova Elementi di Combustibile: Fuel Elements Test) is an experimental fast sodium-cooled reactor with a power of 120 MWt. This reactor aims at studying the behaviour of fuel elements under thermal and neutron conditions comparable with those existing in fast power nuclear facilities. Given the particular structure of the core, the complex operations to be performed in the transfer cell and the strict operating conditions of the central channel, two experimental facilities, CPC-1 and CEDI, have been designed as a support to the construction of the reactor. CPC-1 is a 1:1 scale model of the channel, transfer-cell and loop unit of the channel, whereas CEDI is a sodium-cooled loop which enables to carry out tests of isothermal endurance and thermal shocks on the group of seven forced elements, by simulating the thermo-hydraulic and mechanical conditions existing in the reactor. In this paper some experimental test are briefy discussed and some facilities are listed, both for the CPC-1 and for the CEDI. (Auth.)

  15. Shadow corrosion testing in the INCA facility in the Studsvik R2 reactor

    International Nuclear Information System (INIS)

    Nystrand, A.C.; Lassing, A.

    1999-01-01

    Shadow corrosion is a phenomenon which occurs when zirconium alloys are in contact with or in proximity to other metallic objects in a boiling water reactor environment (BWR, RBMK, SGHWR etc.). An enhanced corrosion occurs on the zirconium alloy with the appearance of a 'shadow' of the metallic object. The magnitude of the shadow corrosion can be significant, and is potentially limiting for the lifetime of certain zirconium alloy components in BWRs and other reactors with a similar water chemistry. In order to evaluate the suitability of the In-Core Autoclave (INCA) in the Studsvik R2 materials testing reactor as an experimental facility for studying shadow corrosion, a demonstration test has been performed. A number of test specimens consisting of Zircaloy-2 tubing in contact with Inconel were exposed in an oxidising water chemistry. Some of the specimens were placed within the reactor core and some above the core. The conclusion of this experiment after post irradiation examination is that it is possible to use the INCA facility in the Studsvik R2 reactor to develop a significant level of shadow corrosion after only 800 hours of irradiation. (author)

  16. Reactor operation safety information document

    Energy Technology Data Exchange (ETDEWEB)

    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. PANDA: a Large Scale Multi-Purpose Test Facility for LWR Safety Research

    Energy Technology Data Exchange (ETDEWEB)

    Dreier, Joerg; Paladino, Domenico; Huggenberger, Max; Andreani, Michele [Laboratory for Thermal-Hydraulics, Nuclear Energy and Safety Research Department, Paul Scherrer Institut (PSI), CH-5232 Villigen PSI (Switzerland); Yadigaroglu, George [ETH Zuerich, Technoparkstrasse 1, Einstein 22- CH-8005 Zuerich (Switzerland)

    2008-07-01

    PANDA is a large-scale multi-purpose thermal-hydraulics test facility, built and operated by PSI. Due to its modular structure, PANDA provides flexibility for a variety of applications, ranging from integral containment system investigations, primary system tests, component experiments to large-scale separate-effects tests. For many applications, the experimental results are directly used for example for concept demonstrations or for the characterisation of phenomena or components, but all the experimental data generated in the various test campaigns is unique and was or/and will still be widely used for the validation and improvement of a variety of computer codes, including codes with 3D capabilities, for reactor safety analysis. The paper provides an overview of the already completed and on-going research programs performed in the PANDA facility in the different area of applications, including the main results and conclusions of the investigations. In particular the advanced passive containment cooling system concept investigations of the SBWR, ESBWR as well as of the SWR1000 in relation to various aspects are presented and the main findings are summarised. Finally the goals, planned investigations and expected results of the on-going OECD project SETH-2 are presented. (authors)

  18. PANDA: a Large Scale Multi-Purpose Test Facility for LWR Safety Research

    International Nuclear Information System (INIS)

    Dreier, Joerg; Paladino, Domenico; Huggenberger, Max; Andreani, Michele; Yadigaroglu, George

    2008-01-01

    PANDA is a large-scale multi-purpose thermal-hydraulics test facility, built and operated by PSI. Due to its modular structure, PANDA provides flexibility for a variety of applications, ranging from integral containment system investigations, primary system tests, component experiments to large-scale separate-effects tests. For many applications, the experimental results are directly used for example for concept demonstrations or for the characterisation of phenomena or components, but all the experimental data generated in the various test campaigns is unique and was or/and will still be widely used for the validation and improvement of a variety of computer codes, including codes with 3D capabilities, for reactor safety analysis. The paper provides an overview of the already completed and on-going research programs performed in the PANDA facility in the different area of applications, including the main results and conclusions of the investigations. In particular the advanced passive containment cooling system concept investigations of the SBWR, ESBWR as well as of the SWR1000 in relation to various aspects are presented and the main findings are summarised. Finally the goals, planned investigations and expected results of the on-going OECD project SETH-2 are presented. (authors)

  19. Test of safety injection supply by diesel generator under reactor vessel closed condition

    International Nuclear Information System (INIS)

    Zhang Hao; Bi Fengchuan; Che Junxia; Zhang Jianwen; Yang Bo

    2014-01-01

    The paper studied that the test of diesel generator full load take-up under the condition of actual safety injection and reactor vessel closed in Ningde nuclear project unit l. It is proved that test result accorded with design criteria, meanwhile, the test was removed from the key path of project schedule, which cut a huge cost. (authors)

  20. Radiological controls and worker and public health and safety: An independent safety assessment of Department of Energy nuclear reactor facilities

    International Nuclear Information System (INIS)

    Tew, J.L.; Miles, M.E.; Knuth, D.; Boyd, R.

    1981-02-01

    DOE has formed a Nuclear Facilities Personnel Qualification and Training (NFPQT) Committee to assess the implications of the Report of the President's Commission on the Accident at Three Mile Island that are applicable to DOE's nuclear reactor operations. Thirteen DOE nuclear reactors were reviewed by the Committee. This report was prepared to provide a measure of how the radiological control and environmental practices at the 13 individual DOE reactor facilities measure up to (1) the recommendations contained in the Report of the President's Commission on the Accident at Three Mile Island, (2) the requirements and guidelines contained, and (3) the requirements of the applicable Title and Part of the Code of Federal Regulations

  1. Final safety and hazards analysis for the Battelle LOCA simulation tests in the NRU reactor

    International Nuclear Information System (INIS)

    Axford, D.J.; Martin, I.C.; McAuley, S.J.

    1981-04-01

    This is the final safety and hazards report for the proposed Battelle LOCA simulation tests in NRU. A brief description of equipment test design and operating procedure precedes a safety analysis and hazards review of the project. The hazards review addresses potential equipment failures as well as potential for a metal/water reaction and evaluates the consequences. The operation of the tests as proposed does not present an unacceptable risk to the NRU Reactor, CRNL personnel or members of the public. (author)

  2. Emergency planning and response: An independent safety assessment of Department of Energy nuclear reactor facilities

    International Nuclear Information System (INIS)

    Knuth, D.; Boyd, R.

    1981-02-01

    The Department of Energy (DOE) has formed a Nuclear Facilities Personnel Qualification and Training (NFPQT) Committee to assess the implications of the recommendations contained in the President's Commission Report on the Three Mile Island (TMI) Accident (the Kemeny Commission report) that are applicable to DOE's nuclear reactor operations. Thirteen DOE nuclear reactors have been reviewed. The assessments of the 13 facilities are based on information provided by the individual operator organizations and/or cognizant DOE Field Offices. Additional clarifying information was supplied in some, but not all, instances. This report indicates how these 13 reactor facilities measure up in light of the Kemeny and other TMI-related studies and recommendations, particularly those that have resulted in upgraded Nuclear Regulatory Commission (NRC) requirements in the area of emergency planning and response

  3. Scram and nonlinear reactor system seismic analysis for the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Morrone, A.

    1975-01-01

    A description is given of the analysis and results for the Fast Flux Test Facility (FFTF) reactor system which was analyzed for both scram times and seismic responses such as bending moments and impact forces. The reactor system was represented with a one-dimensional nonlinear mathematical model with two degrees of freedom per node. The results give time history plots of various seismic responses and plots of scram times as a function of control rod travel distance for the most critical scram initiation times. The total scram time considering the effects of the earthquake was still acceptable but about 4 times longer than that calculated without the earthquake. (U.S.)

  4. Design and safety consideration in the High-Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Saito, Shinzo; Tanaka, Toshiuki; Sudo, Yukio; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru

    1990-01-01

    The budget for construction of the High-Temperature Engineering Test Reactor (HTTR) was recently committed by the Government in Japan. The HTTR is a test reactor with thermal output of 30 MW and reactor outlet coolant temperature of 950 deg. C at high temperature test operation. The HTTR plant uses a pin-in-block design core and will be used as an experience leading to high temperature applications. Several major important safety considerations are adopted in the design of the HTTR. These are as follows: 1) A coated particle fuel must not be failed during a normal reactor operation and an anticipated operational occurrence; 2) Two independent and diverse reactor shut-down systems are provided in order to shut down the reactor safely and reliably in any condition; 3) Back-up reactor cooling systems which are safety ones are provided in order to remove residual heat of reactor in any condition; 4) Multiple barriers and countermeasures are provided to contain fission products such as a containment, pressure gradient between the primary and secondary cooling circuit and so on, though coated particle fuels contain fission products with high reliability; 5) The functions of materials used in the primary cooling circuit are separated to be pressure-resisting and heat-resisting in order to resolve material problems and maintain high reliability. The detailed design of the HTTR was completed with extensive accumulation of material data and component tests. (author)

  5. Thermal-Hydraulic Tests for Reactor Core Safety

    International Nuclear Information System (INIS)

    Chun, Se Young; Chung, Moon Ki; Baek, Won Pil and others

    2005-04-01

    The reflood experiments for single rod annulus geometry have been performed to investigate the effect of spacer grid on thermal-hydraulics under reflood conditions. The reflood experimental loop for 6x6 rod bundle with a spacer grid developed in Korea has been provided. About 8000 data points for Post-CHF heat transfer have been obtained from the experiments About 1400 CHF data points for 3x3 Water and 5x5 Freon rod bundles have been obtained. The existing evaluation methodology for core safety under return-to-power conditions has been investigated using KAERI low flow CHF database. The hydraulic tests for turbulence mixing characteristics in subchannel of 5x5 rod bundle have been carried out using advanced measurement technique, LVD and the database for various spacer grids have been provided. In order to measure the turbulence mixing characteristics in details, the hydraulic loop with a magnified 5x5 rod bundle has been prepared. The database which was constructed through a systematic thermal hydraulic tests for the reflood phenomenon, CHF, Post-CHF is surely to be useful to the industry field, the regulation body and the development of thermal-hydraulic analysis code

  6. High flux materials testing reactor HFR Petten. Characteristics of facilities and standard irradiation devices

    International Nuclear Information System (INIS)

    Roettger, H.; Hardt, P. von der; Tas, A.; Voorbraak, W.P.

    1981-01-01

    For the materials testing reactor HFR some characteristic information is presented. Besides the nuclear data for the experiment positions short descriptions are given of the most important standard facilities for material irradiation and radionuclide production. One paragraph deals with the experimental set-ups for solid state and nuclear structure investigations. The information in this report refers to a core type, which is operational since March 1977. The numerical data compiled have been up-dated to January 1981

  7. Thermohydraulic tests in the area of reactor safety done in CDTN

    International Nuclear Information System (INIS)

    Ladeira, L.C.D.

    1990-01-01

    The main experimental works performed in the last five years at the Thermohydraulics Laboratory of the Nuclear Technology Development Center, in the field of reactor safety are briefly described. This paper cover the performing and analysis of pressure drop, heat transfer and mixing tests in 3X3 rod bundle and rewetting tests in single tube section. (autor) [pt

  8. The advanced test reactor national scientific user facility advancing nuclear technology

    International Nuclear Information System (INIS)

    Allen, T.R.; Thelen, M.C.; Meyer, M.K.; Marshall, F.M.; Foster, J.; Benson, J.B.

    2009-01-01

    To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty-student team

  9. The Advanced Test Reactor National Scientific User Facility Advancing Nuclear Technology

    International Nuclear Information System (INIS)

    Allen, T.R.; Benson, J.B.; Foster, J.A.; Marshall, F.M.; Meyer, M.K.; Thelen, M.C.

    2009-01-01

    To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty-student team

  10. Study on personnel qualification for non-destructive tests in the field of reactor safety

    International Nuclear Information System (INIS)

    Trusch, K.; Wuestenberg, H.

    1977-01-01

    The training system for non-destructive testing is described, and the available and necessary personnel is analyzed; the personnel required for reactor safety problems is treated separately. On this basis, the subjects discussed in the study - available personnel, personnel requirements, training, training requirements, and suggestions for realisation - are treated in a general manner to begin with and afterwards with a view to specific problems of reactor safety. The methods employed are adapted to this situation. To obtain the necessary empirical data, questionnaires were set up and distributed, and experts in selected business companies and institutions were interviewed who work in the field of reactor safety or do same training in non-destructive testing. (orig.) [de

  11. Radiological considerations of the reactor cover gas processing system at the FFTF [Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Prevo, P.R.

    1986-09-01

    Radiological and environmental protection experience associated with the reactor cover gas processing system at the Fast Flux Test Facility (FFTF) has been excellent. Personnel radiation exposures received from operating and maintaining the reactor cover gas processing system have been very low, the system has remained free of radioactive particulate contamination through the first seven operating cycles (cesium contamination was detected at the end of Cycle 8A), and releases of radioactivity to the environment have been very low, well below environmental standards. This report discusses these three aspects of fast reactor cover gas purification over the first eight operating cycles of the FFTF (a duration of a little more than four years, from April 1982 through July 1986)

  12. Auditable Safety Analysis and Final Hazard Classification for the 105-N Reactor Zone and 109-N Steam Generator Zone Facility

    International Nuclear Information System (INIS)

    Kloster, G.L.

    1998-07-01

    This document is a graded auditable safety analysis (ASA) and final hazard classification (FHC) for the Reactor/Steam Generator Zone Segment. The Reactor/Steam Generator Zone Segment, part of the N Reactor Complex, that is also known as the Reactor Building and Steam Generator Cells. The installation of the modifications described within to support surveillance and maintenance activities are to be completed by July 1, 1999. The surveillance and maintenance activities addressed within are assumed to continue for the next 15- 20 years, until the initiation of facility D ampersand D (i.e., Interim Safe Storage). The graded ASA in this document is in accordance with EDPI-4.30-01, Rev. 1, Safety Analysis Documentation, (BHI-DE-1) and is consistent with guidance provided by the U.S. Department of Energy. This ASA describes the hazards within the facility and evaluates the adequacy of the measures taken to reduce, control, or mitigate the identified hazards. This document also serves as the FHC for the Reactor/Steam Generator Zone Segment. This FHC is developed through the use of bounding accident analyses that envelope the potential exposures to personnel

  13. Safety assessment of the SMART design during SBLOCA tests using the high pressure safety injection pump of the SMART-ITL facility

    International Nuclear Information System (INIS)

    Bae, Hwang; Ryu, Sung Uk; Jeon, Byong-Guk; Yang, Jin-Hwa; Yoon, Eun-Koo; Shin, Yong-Cheol; Min, Kyoung-Ho; Park, Jong-Kuk; Choi, Nam-Hyun; Bang, Yun-Gon; Seo, Chan-Jong; Yi, Sung-Jae; Park, Hyun-Sik

    2016-01-01

    SMART is a small-sized integral pressurized light water reactor designed by the Korea Atomic Energy Research Institute (KAERI) from 1997 and received standard design approval (SDA) by the Korean regulatory body in July 2012. Single reactor pressure vessel contains all of the main components including a pressurizer (PZR), steam generators (SG) and reactor coolant pumps (RCP) without any large-size pipes. Several tests to verify a safety and performance of SMART design were carried out. This paper introduces a comparison with three SBLOCA tests. Overall thermal-hydraulic phenomena were observed and showed a traditional trend to decrease a system pressure and temperature. A collapsed water level of the hot side indicated that the safety injection system was successfully operated to recover the reactor coolant system (RCS) and protect the core uncover. An SBLOCA test simulating a guillotine break on the SIS, SCS, and PSV was performed. It was enough to keep a steady-state condition before the SBLOCA test begins. An actuation signal as the boundary condition was properly simulated during the transient test. The scenarios of the SBLOCA in the SMART design were reproduced well using the SMART-ITL facility. The safety injection is effective to protect the core uncover as well as to cool down the RCS. All of the measured parameters show reasonable behaviors

  14. Safety assessment of the SMART design during SBLOCA tests using the high pressure safety injection pump of the SMART-ITL facility

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Hwang; Ryu, Sung Uk; Jeon, Byong-Guk; Yang, Jin-Hwa; Yoon, Eun-Koo; Shin, Yong-Cheol; Min, Kyoung-Ho; Park, Jong-Kuk; Choi, Nam-Hyun; Bang, Yun-Gon; Seo, Chan-Jong; Yi, Sung-Jae; Park, Hyun-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    SMART is a small-sized integral pressurized light water reactor designed by the Korea Atomic Energy Research Institute (KAERI) from 1997 and received standard design approval (SDA) by the Korean regulatory body in July 2012. Single reactor pressure vessel contains all of the main components including a pressurizer (PZR), steam generators (SG) and reactor coolant pumps (RCP) without any large-size pipes. Several tests to verify a safety and performance of SMART design were carried out. This paper introduces a comparison with three SBLOCA tests. Overall thermal-hydraulic phenomena were observed and showed a traditional trend to decrease a system pressure and temperature. A collapsed water level of the hot side indicated that the safety injection system was successfully operated to recover the reactor coolant system (RCS) and protect the core uncover. An SBLOCA test simulating a guillotine break on the SIS, SCS, and PSV was performed. It was enough to keep a steady-state condition before the SBLOCA test begins. An actuation signal as the boundary condition was properly simulated during the transient test. The scenarios of the SBLOCA in the SMART design were reproduced well using the SMART-ITL facility. The safety injection is effective to protect the core uncover as well as to cool down the RCS. All of the measured parameters show reasonable behaviors.

  15. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2010-03-01

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these

  16. Preliminary reactor physics calculations for Exxon LWR fuel testing in the power burst facility

    International Nuclear Information System (INIS)

    Olson, W.O.; Nigg, D.W.

    1981-05-01

    The PFB reactor is being considered as an irradiation facility to test LWR fuel rods for Exxon Nuclear Company. Requested test conditions are 18 kW/ft axial peak steady state power in 2.5% initial enrichment, 20,000 MWd/Tu exposed rods. Multigroup transport theory calculations (S/sub n/ and Monte Carlo) showed that this was unattainable in the standard PBF test loop. Thus, a flux multiplier was developed in the form of a Zr-2-clad 0.15-inch thick cylindrical shell of 35% enriched, 88% T.D. UO 2 replacing the flow divider, surrounding the rod within the in-pile tube in PFB. With this flux multiplier installed and assuming an average water density of 0.86 g/cm 3 within the test loop, a Figure of Merit (FOM) for a single-rod test assembly of 0.86 kW/ft-MW +- 5% (at 95% confidence level) was calculated. This FOM is the axial peak linear test rod power per megawatt of reactor power. A reactor power of about 21 megawatts will therefore be required to supply the requested linear test rod axial peak heating rate of 18 kW/ft

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

    International Nuclear Information System (INIS)

    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

  18. Operational safety assessment of underground test facilities for mined geologic waste disposal

    International Nuclear Information System (INIS)

    Elder, H.K.

    1993-01-01

    This paper describes the operational safety assessment for the underground facilities for the exploratory studies facility (ESF) at the Yucca Mountain Project. The systematic identification and evaluation of hazards related to the ESF is an integral part of the systems engineering process; whereby safety is considered during planning, design, testing, and construction. A largely qualitative approach based on the analysis of potential accidents was used since radiological safety analysis was not required. The risk assessment summarized credible accident scenarios and the design provides mitigation of the risks to a level that the facility can be constructed and operated with an adequate level of safety. The risk assessment also provides reasonable assurance that all identifiable major accident scenarios have been reviewed and design mitigation features provided to ensure an adequate level of safety

  19. Area Safety Program for the tokamak fusion test reactor (TFTR)

    International Nuclear Information System (INIS)

    Rappe, G.M.

    1984-10-01

    Overall the Area Safety Program has proved to be a very successful operation. There is no doubt that a safety program organized through line management is the best way to involve all personnel. Naturally, when the program was first started, there was some criticism and a certain resistance on the part of a few individuals to fully participate. However, once the program was underway and it could be seen that it was working to everyone's advantage, this reluctance disappeared and a spirit of full cooperation is now enjoyed. It is very important that for this success to continue there must be a two way flow of information, both from the Area Safety Coordinators up through line management, and from senior management, with decisions and answers, back down through the management chain with the utmost dispatch. As with all programs, there is still room for improvement. This program has started a review cycle with a view to streamlining certain areas and possibly increasing its scope in others

  20. Safety of Basic nuclear facilities (INB) other than electronuclear reactors. Lessons learned from declared significant events in 2011 and 2012

    International Nuclear Information System (INIS)

    2013-01-01

    The first part of this report presents the different types of basic nuclear facilities other than electronuclear reactors. These installations can be industrial installations dedicated or not to the nuclear fuel cycle, research and support installations, be definitively stopped or being dismantled, or radioactive waste storage installations. After a comment of the main trends noticed in 2011 and 2012, the report proposes a transverse analysis of events which occurred in these installations. These events are related to various risks: dissemination of radioactive materials, exposure to ionizing radiations, criticality, fire and explosion, handling operations, loss of electric supplies or fluids, external aggression. Other events are those significant for the environment with a radiological component, or related to periodic controls and tests. The causes of these events are analysed. Specific events are presented which occurred on different sites (in the MELOX plant, in Areva sites in La Hague, Pierrelatte, in CEA sites in Cadarache and Saclay, in a fuel factory in Romans). Other topics are finally addressed: safety measures after the Fukushima accident, safety and radiation protection management systems of Areva and CEA, dismantling of nuclear installations

  1. Technical assessment: An independent safety assessment of Department of Energy nuclear reactor facilities

    International Nuclear Information System (INIS)

    Brodsky, R.S.

    1981-02-01

    Inherent in the design of DOE reactors under review are many features which provide significant protection against the likelihood of TMI-type accidents. In addition, other features in the design or operating characteristics would tend to limit or reduce the consequences of the accident. Some of these features were discussed earlier in this report. However, some of the events included within the TMI accident sequence contain technical implications for the DOE reactors. These implications were reviewed by this Assessment Team, and the results of this review are reported in this and the following sections of this report. It is also important to reemphasize that as a result of this review, no major TMI-related safety issues have been identified that would indicate that these DOE reactors cannot be operated in a safe manner. Rather, the findings of this report, by nature, generally reemphasize and support ongoing DOE efforts and identify areas for additional improvements

  2. Safety-licensing assessment of NASAP reactor concepts and fuel cycle facilities

    International Nuclear Information System (INIS)

    Lipinski, W.C.; Prohammer, F.G.; van Erp, J.B.; Seefeldt, W.B.

    1978-06-01

    Assessments are presented of the safety/licensability of reactor concepts based on information supplied by the Nonproliferation Alternative Systems Assessment Program (NASAP) characterization contractors in their updated responses to the data package for NASAP Rolling Report II. The assessment of the LMFBR includes information from a characterization contractor on alternate fuel cycles but does not include information provided by a characterization contractor on plant-related safety issues. The information provided by the characterization contractors was supplemented by assessments provided by the U. S. Nuclear Regulatory Commission

  3. Creep-Fatigue Damage Evaluation of a Model Reactor Vessel and Reactor Internals of Sodium Test Facility according to ASME-NH and RCC-MRx Codes

    International Nuclear Information System (INIS)

    Lim, Dong-Won; Lee, Hyeong-Yeon; Eoh, Jae-Hyuk; Son, Seok-Kwon; Kim, Jong-Bum; Jeong, Ji-Young

    2016-01-01

    The objective of the STELLA-2 is to support the specific design approval for PGSFR by synthetic reviews of key safety issues and code validations through the integral effect tests. Due to its high temperature operation in SFRs (and in a testing facility) up to 550 °C, thermally induced creep-fatigue damage is very likely in components including a reactor vessel, reactor internals (interior structures), heat exchangers, pipelines, etc. In this study, structural integrity of the components such as reactor vessel and internals in STELLA-2 has been evaluated against creep-fatigue failures at a concept-design step. As 2D analysis yields far conservative results, a realistic 3D simulation is performed by a commercial software. A design integrity guarding against a creep-fatigue damage failure operating at high temperature was evaluated for the reactor vessel with its internal structure of the STELLA-2. Both the high temperature design codes were used for the evaluation, and results were compared. All the results showed the vessel as a whole is safely designed at the given operating conditions, while the ASME-NH gives a conservative evaluation

  4. Creep-Fatigue Damage Evaluation of a Model Reactor Vessel and Reactor Internals of Sodium Test Facility according to ASME-NH and RCC-MRx Codes

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Dong-Won; Lee, Hyeong-Yeon; Eoh, Jae-Hyuk; Son, Seok-Kwon; Kim, Jong-Bum; Jeong, Ji-Young [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The objective of the STELLA-2 is to support the specific design approval for PGSFR by synthetic reviews of key safety issues and code validations through the integral effect tests. Due to its high temperature operation in SFRs (and in a testing facility) up to 550 °C, thermally induced creep-fatigue damage is very likely in components including a reactor vessel, reactor internals (interior structures), heat exchangers, pipelines, etc. In this study, structural integrity of the components such as reactor vessel and internals in STELLA-2 has been evaluated against creep-fatigue failures at a concept-design step. As 2D analysis yields far conservative results, a realistic 3D simulation is performed by a commercial software. A design integrity guarding against a creep-fatigue damage failure operating at high temperature was evaluated for the reactor vessel with its internal structure of the STELLA-2. Both the high temperature design codes were used for the evaluation, and results were compared. All the results showed the vessel as a whole is safely designed at the given operating conditions, while the ASME-NH gives a conservative evaluation.

  5. Development of a methodology for safety classification on a non-reactor nuclear facility illustrated using an specific example

    International Nuclear Information System (INIS)

    Scheuermann, F.; Lehradt, O.; Traichel, A.

    2015-01-01

    To realize the safety of personnel and environment systems and components of nuclear facilities are classified according to their potential danger into safety classes. Based on this classification different demands on the manufacturing quality result. The objective of this work is to present the standardized method developed by NUKEM Technologies Engineering Services for the categorization into the safety classes restricted to Non-reactor nuclear facilities (NRNF). Exemplary the methodology is used on the complex Russian normative system (four safety classes). For NRNF only the lower two safety classes are relevant. The classification into the lowest safety class 4 is accordingly if the maximum resulting dose following from clean-up actions in case of incidents/accidents remains below 20 mSv and the volume activity restrictions of set in NRB-99/2009 are met. The methodology is illustrated using an example. In short the methodology consists of: - Determination of the working time to remove consequences of incidents, - Calculation of the dose resulting from direct radiation and due to inhalation during these works. The application of this methodology avoids over-conservative approaches. As a result some previously higher classified equipment can be classified into the lower safety class.

  6. Mirror Fusion Test Facility: an intermediate device to a mirror fusion reactor

    International Nuclear Information System (INIS)

    Karpenko, V.N.

    1983-01-01

    The Mirror Fusion Test Facility (MFTF-B) now under construction at Lawrence Livermore National Laboratory represents more than an order-of-magnitude step from earlier magnetic-mirror experiments toward a future mirror fusion reactor. In fact, when the device begins operating in 1986, the Lawson criteria of ntau = 10 14 cm -3 .s will almost be achieved for D-T equivalent operation, thus signifying scientific breakeven. Major steps have been taken to develop MFTF-B technologies for tandem mirrors. Steady-state, high-field, superconducting magnets at reactor-revelant scales are used in the machine. The 30-s beam pulses, ECRH, and ICRH will also introduce steady-state technologies in those systems

  7. Reactor safety research in Sweden

    International Nuclear Information System (INIS)

    Pershagen, B.

    1980-02-01

    Objectives, means and results of Swedish light water reactor safety research during the 1970s are reviewed. The expenditure is about 40 Million Swkr per year excluding industry. Large efforts have been devoted to experimental studies of loss of coolant accidents. Large scale containment response tests for simulated pipe breaks were carried out at the Marviken facility. At Studsvik a method for testing fuel during fast power changes has been developed. Stress corrosion, crack growth and the effect of irradiation on the strength ductility of Zircaloy tube was studied. A method for determining the fracture toughness of pressure vessel steel was developed and it was shown that the fracture toughness was lower than earlier assumed. The release of fission products to reactor water was studied and so was the release, transport and removal of airborne radioactive matter for Swedish BWRs and PWRs. Test methods for iodine filter systems were developed. A system for continuous monitoring of radioactive noble gas stack release was developed for the Ringhals plant. Attention was drawn to the importance of the human factor for reactor safety. Probabilistic methods for risk analysis were applied to the Barsebaeck 2 and Forsmark 3 boiling water reactors. Procedures and working conditions for operator personnel were investigated. (GBn)

  8. JRR-3 cold neutron source facility H2-O2 explosion safety proof testing

    International Nuclear Information System (INIS)

    Hibi, T.; Fuse, H.; Takahashi, H.; Akutsu, C.; Kumai, T.; Kawabata, Y.

    1990-01-01

    A cold Neutron Source (CNS) will be installed in Japan Research Reactor-3 (JRR-3) in Japan Atomic Energy Research Institute (JAERI) during its remodeling project. This CNS holds liquid hydrogen at a temperature of about 20 K as a cold neutron source moderator in the heavy water area of the reactor to moderate thermal neutrons from the reactor to cold neutrons of about 5 meV energy. In the hydrogen circuit of the CNS safety measures are taken to prevent oxygen/hydrogen reaction (H 2 -O 2 explosion). It is also designed in such manner that, should an H 2 -O 2 explosion take place, the soundness of all the components can be maintained so as not to harm the reactor safety. A test hydrogen circuit identical to that of the CNS (real components designed by TECHNICATOME of France) was manufactured to conduct the H 2 -O 2 explosion test. In this test, the detonation that is the severest phenomenon of the oxygen/hydrogen reaction took place in the test hydrogen circuit to measure the exerted pressure on the components and their strain, deformation, leakage, cracking, etc. Based on the results of this measurement, the structural strength of the test hydrogen circuit was analyzed. The results of this test show that the hydrogen circuit components have sufficient structural strength to withstand an oxygen/hydrogen reaction

  9. Calculational framework for safety analyses of non-reactor nuclear facilities

    International Nuclear Information System (INIS)

    Coleman, J.R.

    1994-01-01

    A calculational framework for the consequences analysis of non-reactor nuclear facilities is presented. The analysis framework starts with accident scenarios which are developed through a traditional hazard analysis and continues with a probabilistic framework for the consequences analysis. The framework encourages the use of response continua derived from engineering judgment and traditional deterministic engineering analyses. The general approach consists of dividing the overall problem into a series of interrelated analysis cells and then devising Markov chain like probability transition matrices for each of the cells. An advantage of this division of the problem is that intermediate output (as probability state vectors) are generated at each calculational interface. The series of analyses when combined yield risk analysis output. The analysis approach is illustrated through application to two non-reactor nuclear analyses: the Ulysses Space Mission, and a hydrogen burn in the Hanford waste storage tanks

  10. Test facility for auxiliary cooling system (ACS) of fast breeder reactor for Power Reactor and Nuclear Fuel Development Corporation (PNC)

    International Nuclear Information System (INIS)

    1983-01-01

    In preparation of constructing ''Monju'', a prototype fast breeder reactor, PNC has been pushing forward its research and development projects and the ACS was constructed under these projects. The auxiliary cooling system is an important engineered safety feature, and is used for safe removal of heat from the reactor at the shutdown. The ACS serves as a means of testing and assessing the auxiliary cooling system for the ''Monju'' and is designed and manufactured to have one fifth capacity of the Monju. The air heat exchanger and the ACS system was designed to withstand higher temperature range of the conventional design code (MITI-501), and finned tubes were applied for effective heat removal. Preheating system was designed to heat up the whole system over 200 0 C within 20 hours to prevent sodium from freezing. Basic performance of ACS was verified satisfactorily by a series of performance tests, such as start up test, flow rate measurement and preheating test before delivery. The experience from designing and construction of ACS and data obtained by these tests will be very instructive for designing and construction of the ''Monju''. (author)

  11. STEADY STATE MODELING OF THE MINIMUM CRITICAL CORE OF THE TRANSIENT REACTOR TEST FACILITY

    Energy Technology Data Exchange (ETDEWEB)

    Anthony L. Alberti; Todd S. Palmer; Javier Ortensi; Mark D. DeHart

    2016-05-01

    With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. The DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum, air-cooled, nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific scenarios range from simple temperature transients to full fuel melt accidents. DOE has expressed a desire to develop a simulation capability that will accurately model the experiments before they are irradiated at the facility. It is the aim for this capability to have an emphasis on effective and safe operation while minimizing experimental time and cost. The multi physics platform MOOSE has been selected as the framework for this project. The goals for this work are to identify the fundamental neutronics properties of TREAT and to develop an accurate steady state model for future multiphysics transient simulations. In order to minimize computational cost, the effect of spatial homogenization and angular discretization are investigated. It was found that significant anisotropy is present in TREAT assemblies and to capture this effect, explicit modeling of cooling channels and inter-element gaps is necessary. For this modeling scheme, single element calculations at 293 K gave power distributions with a root mean square difference of 0.076% from those of reference SERPENT calculations. The minimum critical core configuration with identical gap and channel treatment at 293 K resulted in a root mean square, total core, radial power distribution 2.423% different than those of reference SERPENT solutions.

  12. Capsule safety analysis of PRTF irradiation facility

    International Nuclear Information System (INIS)

    Suwarto

    2013-01-01

    Power Ramp Test Facility (PRTF) is an irradiation facility used for fuel testing of power reactor. PRTF has a capsule which is a test fuel rod container. During operation, pressurized water of 160 bars flows through in the capsule. Due to the high pressure it should be analyzed the impact of the capsule on reactor core safety. This analysis has purpose to calculate the ability of capsule pressure capacity. The analysis was carried out by calculating pressure capacity. From the calculating results it can be concluded that the capsule with pressure capacity of 438 bars will be safe to prevent the operation pressure of PRTF. (author)

  13. Establishment and prioritization of relevant factors to the safety of fuel cycle facilities non reactor through dynamics archetypes evaluation

    International Nuclear Information System (INIS)

    Sousa, Anna Leticia Barbosa de

    2012-01-01

    The present work aims to establish and prioritize factors that are important to the safety of nuclear fuel cycle facilities in order to model, analyze and design safety as a physical system, employing systemic models in an innovative way. This work takes into consideration the fact that models that use adaptations of methodologies for nuclear reactors will not properly work due to the specificities of fuel cycle facilities. Based on the fundamentals of the theory of systems, the four levels of system thinking, and the relationship of eight socio technical factors, a mental model has been developed for safety management in the nuclear fuel cycle context. From this conceptual model, safety archetypes were constructed in order to identify and highlight the processes of change and decision making that allow the system to migrate to a state of loss of safety. After that, stock and flow diagrams were created so that their behavior could be assessed by the system's dynamics. The results from the analysis using the model that simulates the dynamic behavior of the variables (socio technical factors) indicated, as expected, that the system's dynamics proved to be an appropriate and efficient tool for modeling fuel cycle safety as an emergent property. (author)

  14. Gas Cooled Fast Breeder Reactor cost estimate for a circulator test facility (modified HTGR circulator test facility)

    International Nuclear Information System (INIS)

    1979-10-01

    This is a conceptual design cost estimate for a Helium Circulator Test Facility to be located at the General Atomic Company, San Diego, California. The circulator, drive motors, controllers, thermal barrier, and circulator service module installation costs are part of the construction cost included

  15. Reactor safety impact of functional test intervals: an application of Bayesian decision theory

    International Nuclear Information System (INIS)

    Buoni, F.B.

    1978-01-01

    Functional test intervals for important nuclear reactor systems can be obtained by viewing safety assessment as a decision process and functional testing as a Bayesian learning or information process. A preposterior analysis is used as the analytical model to find the preposterior expected reliability of a system as a function of test intervals. Persistent and transitory failure models are shown to yield different results. Functional tests of systems subject to persistent failure are effective in maintaining system reliability goals. Functional testing is not effective for systems subject to transitory failure; preventive maintenance must be used. A Bayesian posterior analysis of testing data can discriminate between persistent and transitory failure. The role of functional testing is seen to be an aid in assessing the future performance of reactor systems

  16. Introduction to the modified TROI test facility for fuel coolant interaction under a submerged reactor vessel

    International Nuclear Information System (INIS)

    Na, Young Su; Hong, Seong-Wan; Song, Jin Ho; Hong, Seong-Ho

    2014-01-01

    The molten Fuel-Coolant Interaction (FCI) can threaten the integrity of the reactor cavity under a severe accident. A steam explosion can be occurred by the rapid energy transfer in the high-temperature corium melt jet penetrating into water, which makes the dynamic load applying to the surrounding structure. Before a steam explosion, the corium melt jet breaks into small-sized particles, and the steam is generated continuously by the film boiling on the hot surface of the melt contacting with water. The premixing phase consisting of the corium melt, water, and steam can determine the intensity of the steam explosion. Unfortunately, the previous experimental studies on the FCI phenomena have carried out under a free fall of the corium melt jet in a gas phase before interacting with water. The previous TROI (Test for Real cOrium Interaction with water) test facility, that is a well-known test facility for the FCI phenomena in the world, has observed a steam explosion under a free fall of a corium melt jet in a gas phase before contacting a coolant since 2000, which is changing to simulate the FCI phenomena under a submerged reactor vessel. This study introduces the modified TROI test facility as shown in Fig. 1 and the considerations for the experiment with success. The previous TROI test facility, that has observed the molten Fuel-Coolant Interaction (FCI) with a free fall of the prototypic corium melt in a gas phase before contacting a coolant, was modified to simulate the FCI phenomena under a submerged reactor vessel for the assessment of the In-Vessel Retention (IVR) concept, i.e., without a free-fall distance of the corium melt before contacting water. The superheated prototypic corium melt created by the cold crucible melting method moves on a releasing valve newly installed just above the water level in the interaction vessel. The corium melt will stay on a releasing valve in less than 0.2 seconds to reduce heat loss for preventing the solidification, and

  17. Testing and qualification of Control and Safety Rod and its drive mechanism of Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Rajan Babu, V.; Veerasamy, R.; Patri, Sudheer; Ignatius Sundar Raj, S.; Kumar Krovvidi, S.C.S.P.; Dash, S.K.; Meikandamurthy, C.; Rajan, K.K.; Puthiyavinayagam, P.; Chellapandi, P.; Vaidyanathan, G.; Chetal, S.C.

    2010-01-01

    Prototype Fast Breeder Reactor (PFBR) has two independent fast acting diverse shutdown systems. The absorber rod of the first system is called Control and Safety Rod (CSR). CSR and its Drive Mechanism (CSRDM) are used for reactor control and for safe shutdown of the reactor by scram action. In view of the safety role, the qualification of CSRDM is one of the important requirements. CSR and CSRDM were qualified in two stages by extensive testing. In the first stage, the critical subassemblies of the mechanism, such as scram release electromagnet, hydraulic dashpot and dynamic seals and CSR subassembly, were tested and qualified individually simulating the operating conditions of the reactor. Experiments were also carried out on sodium vapour deposition in the annular gaps between the stationary and mobile parts of the mechanism. In the second stage, full-scale CSRDM and CSR were subjected to all the integrated functional tests in air, hot argon and subsequently in sodium simulating the operating conditions of the reactor and finally subjected to endurance tests. Since the damage occurring in CSRDM and CSR is mainly due to fatigue cycles during scram actions, the number of test cycles was decided based on the guidelines given in ASME, Section III, Div. 1. The results show that the performance of CSRDM and CSR is satisfactory. Subsequent to the testing in sodium, the assemblies having contact with liquid sodium/sodium vapour were cleaned using CO 2 process and the total cleaning process has been established, so that the mechanism can be reused in sodium. The various stages of qualification programmes have raised the confidence level on the performance of the system as a whole for the intended and reliable operation in the reactor.

  18. Local flow distribution analysis inside the reactor pools of KALIMER-600 and PDRC performance test facility

    International Nuclear Information System (INIS)

    Jeong, Ji Hwan; Hwang, Seong Won; Choi, Kyeong Sik

    2010-05-01

    In the study, 3-dimensional thermal hydraulic analysis was carried out focusing on the thermal hydraulic behavior inside the reactor pools for both KALIMER-600 and one-fifth scale-down test facility. STAR-CD, one of the commercial CFD codes, was used to analyze 3-dimensional incompressible steady-state thermal hydraulic behavior in both designs of KALIMER-600 and the scale-down test facility. In the KALIMER-600 CFD analysis, the pressure drops in the core and IHX gave a good agreement within 1% error range. It was found that the porous media model was appropriate to analyze the pressure distribution inside reactor core and IHX. Also, a validation analysis showed the pressure drop through the porous media under the condition of 80% flow rate and thermal power was calculated 64% less than in 100% condition giving a physically reasonable analytic result. Since the temperatures in the hot-side pool and cold-side pool were estimated to be very close to 540 and 390 .deg. C specified on the design values respectively, the CFD models of heat source and sink was confirmed. Through the study, the methodology of 3-dimensional CFD analysis about KALIMER-600 has been established and proven. Performed with the methodology, the analysis data such as flow velocity, temperature and pressure distribution were compared by normalizing those data for the actual sized modeling and scale-down modeling. As a result, the characteristics of thermal hydraulic behavior were almost identical for the actual sized modeling and scale-down modeling and the similarity scaling law used in the design of the sodium test facility by KAERI was found to be correct

  19. Performance Test of Korea Heat Load Test Facility (KoHLT-EB) for the Plasma Facing Components of Fusion Reactor

    International Nuclear Information System (INIS)

    Kim, Suk-Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae-Sung; Lee, Dong Won; Cho, Seungyon

    2014-01-01

    The main components of the plasma facing components (PFCs) in the tokamak are the blanket first wall and divertor, which include the armour materials, the heat sink with the cooling mechanism, and the diagnostics devices for the temperature measurement. The Korea Heat Load Test facility by using electron beam (KoHLT-EB) has been operating for the plasma facing components to develop fusion engineering. This electron beam facility was constructed using a 300 kW electron gun and a cylindrical vacuum chamber. Performance tests were carried out for the calorimetric calibrations with Cu dummy mockup and for the heat load test of large Cu module. For the simulation of the heat load test of each mockup, the preliminary thermal-hydraulic analyses with ANSYS-CFX were performed. For the development of the plasma facing components in the fusion reactors, test mockups were fabricated and tested in the high heat flux test facility. To perform a beam profile test, an assessment of the possibility of electron beam Gaussian power density profile and the results of the absorbed power for that profile before the test starts are needed. To assess the possibility of a Gaussian profile, for the qualification test of the Gaussian heat load profile, a calorimeter mockup and large Cu module were manufactured to simulate real heat. For this high-heat flux test, the Korean high-heat flux test facility using an electron beam system was constructed. In this facility, a cyclic heat flux test will be performed to measure the surface heat flux, surface temperature profile, and cooling capacity

  20. Performance Test of Korea Heat Load Test Facility (KoHLT-EB) for the Plasma Facing Components of Fusion Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Suk-Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae-Sung; Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The main components of the plasma facing components (PFCs) in the tokamak are the blanket first wall and divertor, which include the armour materials, the heat sink with the cooling mechanism, and the diagnostics devices for the temperature measurement. The Korea Heat Load Test facility by using electron beam (KoHLT-EB) has been operating for the plasma facing components to develop fusion engineering. This electron beam facility was constructed using a 300 kW electron gun and a cylindrical vacuum chamber. Performance tests were carried out for the calorimetric calibrations with Cu dummy mockup and for the heat load test of large Cu module. For the simulation of the heat load test of each mockup, the preliminary thermal-hydraulic analyses with ANSYS-CFX were performed. For the development of the plasma facing components in the fusion reactors, test mockups were fabricated and tested in the high heat flux test facility. To perform a beam profile test, an assessment of the possibility of electron beam Gaussian power density profile and the results of the absorbed power for that profile before the test starts are needed. To assess the possibility of a Gaussian profile, for the qualification test of the Gaussian heat load profile, a calorimeter mockup and large Cu module were manufactured to simulate real heat. For this high-heat flux test, the Korean high-heat flux test facility using an electron beam system was constructed. In this facility, a cyclic heat flux test will be performed to measure the surface heat flux, surface temperature profile, and cooling capacity.

  1. Safety analysis for research reactors

    International Nuclear Information System (INIS)

    2008-01-01

    The aim of safety analysis for research reactors is to establish and confirm the design basis for items important to safety using appropriate analytical tools. The design, manufacture, construction and commissioning should be integrated with the safety analysis to ensure that the design intent has been incorporated into the as-built reactor. Safety analysis assesses the performance of the reactor against a broad range of operating conditions, postulated initiating events and other circumstances, in order to obtain a complete understanding of how the reactor is expected to perform in these situations. Safety analysis demonstrates that the reactor can be kept within the safety operating regimes established by the designer and approved by the regulatory body. This analysis can also be used as appropriate in the development of operating procedures, periodic testing and inspection programmes, proposals for modifications and experiments and emergency planning. The IAEA Safety Requirements publication on the Safety of Research Reactors states that the scope of safety analysis is required to include analysis of event sequences and evaluation of the consequences of the postulated initiating events and comparison of the results of the analysis with radiological acceptance criteria and design limits. This Safety Report elaborates on the requirements established in IAEA Safety Standards Series No. NS-R-4 on the Safety of Research Reactors, and the guidance given in IAEA Safety Series No. 35-G1, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, providing detailed discussion and examples of related topics. Guidance is given in this report for carrying out safety analyses of research reactors, based on current international good practices. The report covers all the various steps required for a safety analysis; that is, selection of initiating events and acceptance criteria, rules and conventions, types of safety analysis, selection of

  2. Digital System Reliability Test for the Evaluation of safety Critical Software of Digital Reactor Protection System

    Directory of Open Access Journals (Sweden)

    Hyun-Kook Shin

    2006-08-01

    Full Text Available A new Digital Reactor Protection System (DRPS based on VME bus Single Board Computer has been developed by KOPEC to prevent software Common Mode Failure(CMF inside digital system. The new DRPS has been proved to be an effective digital safety system to prevent CMF by Defense-in-Depth and Diversity (DID&D analysis. However, for practical use in Nuclear Power Plants, the performance test and the reliability test are essential for the digital system qualification. In this study, a single channel of DRPS prototype has been manufactured for the evaluation of DRPS capabilities. The integrated functional tests are performed and the system reliability is analyzed and tested. The results of reliability test show that the application software of DRPS has a very high reliability compared with the analog reactor protection systems.

  3. Preliminary design of safety and interlock system for indian test facility of diagnostic neutral beam

    International Nuclear Information System (INIS)

    Tyagi, Himanshu; Soni, Jignesh; Yadav, Ratnakar; Bandyopadhyay, Mainak; Rotti, Chandramouli; Gahlaut, Agrajit; Joshi, Jaydeep; Parmar, Deepak; Bansal, Gourab; Pandya, Kaushal; Chakraborty, Arun

    2016-01-01

    Highlights: • Indian Test Facility being built to characterize DNB for ITER delivery. • Interlock system required to safeguard the investment incurred in building the facility and protecting ITER deliverable components. • Interlock levels upto 3IL-3 identified. • Safety instrumented system for occupational safety being designed. Safety I&C functions of SIL-2 identified. • The systems are based on ITER PIS and PSS design guidelines. - Abstract: Indian Test Facility (INTF) is being built in Institute For Plasma Research to characterize Diagnostic Neutral Beam in co-operation with ITER Organization. INTF is a complex system which consists of several plant systems like beam source, gas feed, vacuum, cryogenics, high voltage power supplies, high power RF generators, mechanical systems and diagnostics systems. Out of these, several INTF components are ITER deliverable, that is, beam source, beam line components and power supplies. To ensure successful operation of INTF involving integrated operation of all the constituent plant systems a matured Data Acquisition and Control System (DACS) is required. The INTF DACS is based on CODAC platform following on PCDH (Plant Control Design Handbook) guidelines. The experimental phases involve application of HV power supplies (100 KV) and High RF power (∼800 KW) which will produce energetic beam of maximum power 6MW within the facility for longer durations. Hence the entire facility will be exposed tohigh heat fluxes and RF radiations. To ensure investment protection and to provide occupational safety for working personnel a matured Safety and Interlock system is required for INTF. The Safety and Interlock systems are high-reliability I&C systems devoted completely to the specific functions. These systems will be separate from the conventional DACS of INTF which will handle the conventional control and acquisition functions. Both, the Safety and Interlock systems are based on IEC 61511 and IEC 61508 standards as

  4. Preliminary design of safety and interlock system for indian test facility of diagnostic neutral beam

    Energy Technology Data Exchange (ETDEWEB)

    Tyagi, Himanshu, E-mail: htyagi@iter-india.org [ITER-India, Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Soni, Jignesh [Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Yadav, Ratnakar; Bandyopadhyay, Mainak; Rotti, Chandramouli [ITER-India, Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Gahlaut, Agrajit [Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Joshi, Jaydeep; Parmar, Deepak [ITER-India, Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India); Bansal, Gourab; Pandya, Kaushal; Chakraborty, Arun [Institute For Plasma Research, Bhat, Gandhinagar, Gujarat (India)

    2016-11-15

    Highlights: • Indian Test Facility being built to characterize DNB for ITER delivery. • Interlock system required to safeguard the investment incurred in building the facility and protecting ITER deliverable components. • Interlock levels upto 3IL-3 identified. • Safety instrumented system for occupational safety being designed. Safety I&C functions of SIL-2 identified. • The systems are based on ITER PIS and PSS design guidelines. - Abstract: Indian Test Facility (INTF) is being built in Institute For Plasma Research to characterize Diagnostic Neutral Beam in co-operation with ITER Organization. INTF is a complex system which consists of several plant systems like beam source, gas feed, vacuum, cryogenics, high voltage power supplies, high power RF generators, mechanical systems and diagnostics systems. Out of these, several INTF components are ITER deliverable, that is, beam source, beam line components and power supplies. To ensure successful operation of INTF involving integrated operation of all the constituent plant systems a matured Data Acquisition and Control System (DACS) is required. The INTF DACS is based on CODAC platform following on PCDH (Plant Control Design Handbook) guidelines. The experimental phases involve application of HV power supplies (100 KV) and High RF power (∼800 KW) which will produce energetic beam of maximum power 6MW within the facility for longer durations. Hence the entire facility will be exposed tohigh heat fluxes and RF radiations. To ensure investment protection and to provide occupational safety for working personnel a matured Safety and Interlock system is required for INTF. The Safety and Interlock systems are high-reliability I&C systems devoted completely to the specific functions. These systems will be separate from the conventional DACS of INTF which will handle the conventional control and acquisition functions. Both, the Safety and Interlock systems are based on IEC 61511 and IEC 61508 standards as

  5. SIMBOL test facility. Safety aspects. Amended version; SIMBOL experimenteerfaciliteit. Veiligheidsaspecten. Geamendeerde versie

    Energy Technology Data Exchange (ETDEWEB)

    Schoonewelle, H.

    1995-12-01

    A detailed description is given of the bubble chamber SIMBOL at the HOR reactor in Delft, Netherlands, including drawings, applied materials and dimensions. Attention is paid to the nuclear and radiological aspects of the installation and an overview is given of the procedures for the realization, the testing and the operational safety of the chamber. Also the results of a failure and accident analysis are presented. Finally, the dismantlement aspects are dealt with. 5 appendices, 12 refs.

  6. Maintenance of reactor safety and control computers at a large government facility

    International Nuclear Information System (INIS)

    Brady, H.G.

    1985-01-01

    In 1950 the US Government contracted the Du Pont Company to design, build, and operate the Savannah River Plant (SRP). At the time, it was the largest construction project ever undertaken by man. It is still the largest of the Department of Energy facilities. In the nearly 35 years that have elapsed, Du Pont has met its commitments to the US Government and set world safety records in the construction and operation of nuclear facilities. Contributing factors in achieving production goals and setting the safety records are a staff of highly qualified personnel, a well maintained plant, and sound maintenance programs. There have been many ''first ever'' achievements at SRP. These ''firsts'' include: (1) computer control of a nuclear rector, and (2) use of computer systems as safety circuits. This presentation discusses the maintenance program provided for these computer systems and all digital systems at SRP. An in-house computer maintenance program that was started in 1966 with five persons has grown to a staff of 40 with investments in computer hardware increasing from $4 million in 1970 to more than $60 million in this decade. 4 figs

  7. Heat Transfer Analysis of the European Pressurized Water Reactor (EPR) Core Catcher Test Facility Volley

    Energy Technology Data Exchange (ETDEWEB)

    Pikkarainen, Mika; Laine, Jani; Purhonen, Heikki; Kyrki-Rajamaeki, Riitta [Lappeenranta University of Technology, P.O. 20 53851 Lappeenranta (Finland); Sairanen, Risto [Radiation and Nuclear Safety Authority, P.O. 14 00881 Helsinki (Finland)

    2008-07-01

    The EPR is designed to cope with severe accidents, involving core meltdown. A specific melt spreading area has been designed within the containment. This core catcher will be flooded by water, which transfers the decay heat to the containment heat removal system. To improve cooling, horizontal flow channels made of cast iron are located also below the core catcher. STUK, the radiation and nuclear safety authority in Finland, wanted an independent study of the functionality of the core catcher design. Effect of the presence of insulation material and boric acid in the cooling water was to be studied, as well as the general behavior of the system in different phases of the flooding of the core melt spreading area. To verify the function of the core catcher design, a scaled down test facility was built at Lappeenranta University of Technology. Since there are some physical restrictions of a test facility computational tools were applied especially for the tests where steady state conditions could not be reached without endangering the integrity of the test facility. This paper introduces the Volley test facility, computational simulations and compares them with the test results. Simulated temperatures of those Volley tests, which could be run until steady state conditions, are very close to the measured temperatures. It can be concluded also, that the temperatures are evidently below the cast iron melting point with heat fluxes used in the tests, if there is a small flow inside the cooling channels or even in case when only a few adjacent cooling channels are totally dry. (authors)

  8. Heat Transfer Analysis of the European Pressurized Water Reactor (EPR) Core Catcher Test Facility Volley

    International Nuclear Information System (INIS)

    Pikkarainen, Mika; Laine, Jani; Purhonen, Heikki; Kyrki-Rajamaeki, Riitta; Sairanen, Risto

    2008-01-01

    The EPR is designed to cope with severe accidents, involving core meltdown. A specific melt spreading area has been designed within the containment. This core catcher will be flooded by water, which transfers the decay heat to the containment heat removal system. To improve cooling, horizontal flow channels made of cast iron are located also below the core catcher. STUK, the radiation and nuclear safety authority in Finland, wanted an independent study of the functionality of the core catcher design. Effect of the presence of insulation material and boric acid in the cooling water was to be studied, as well as the general behavior of the system in different phases of the flooding of the core melt spreading area. To verify the function of the core catcher design, a scaled down test facility was built at Lappeenranta University of Technology. Since there are some physical restrictions of a test facility computational tools were applied especially for the tests where steady state conditions could not be reached without endangering the integrity of the test facility. This paper introduces the Volley test facility, computational simulations and compares them with the test results. Simulated temperatures of those Volley tests, which could be run until steady state conditions, are very close to the measured temperatures. It can be concluded also, that the temperatures are evidently below the cast iron melting point with heat fluxes used in the tests, if there is a small flow inside the cooling channels or even in case when only a few adjacent cooling channels are totally dry. (authors)

  9. Selecting of key safety parameters in reactor nuclear safety supervision

    International Nuclear Information System (INIS)

    He Fan; Yu Hong

    2014-01-01

    The safety parameters indicate the operational states and safety of research reactor are the basis of nuclear safety supervision institution to carry out effective supervision to nuclear facilities. In this paper, the selecting of key safety parameters presented by the research reactor operating unit to National Nuclear Safety Administration that can express the research reactor operational states and safety when operational occurrence or nuclear accident happens, and the interrelationship between them are discussed. Analysis shows that, the key parameters to nuclear safety supervision of research reactor including design limits, operational limits and conditions, safety system settings, safety limits, acceptable limits and emergency action level etc. (authors)

  10. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Appendices

    International Nuclear Information System (INIS)

    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 EMTOMB (entombment). The study results are presented in two volumes. Volume 2 (Appendices) contains the detailed data that support the results given in Volume 1, including unit-component data

  11. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Appendices

    Energy Technology Data Exchange (ETDEWEB)

    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 EMTOMB (entombment). The study results are presented in two volumes. Volume 2 (Appendices) contains the detailed data that support the results given in Volume 1, including unit-component data.

  12. Safety Assessment Document for the Spent Reactor Fuel Geologic Storage Test in the Climax Granite Stock at the Nevada Test site

    International Nuclear Information System (INIS)

    1980-01-01

    The objective of the Spent Fuel Geologic Storage Test in the Climax Granite Stock is to evaluate the response of a granitic rock mass to the underground storage of encapsulated spent reactor fuel in a geometry that simulates a module of a large-scale geologic repository. This document reports an assessment of the safety of conducting this test. Descriptions are provided of the geography, meteorology, hydrology, geology, and seismology of the Climax Site; the effects of postulated natural phenomena and other activities at the nevada Test Site on the safety of the test; and the design and operation of the test facility and associated equipment. Evaluations are made of both the radiological and nonradiological impacts of normal operations, abnormal operations, and postulated accidents. It is concluded that conduct of the spent fuel test at the Climax Site will not result in any undue risk to the public, property, environment, or site employees

  13. Broad-Application Test Reactor

    International Nuclear Information System (INIS)

    Motloch, C.G.

    1992-05-01

    This report is about a new, safe, and operationally efficient DOE reactor of nuclear research and testing proposed for the early to mid- 21st Century. Dubbed the Broad-Application Test Reactor (BATR), the proposed facility incorporates a multiple-application, multiple-mission design to support DOE programs such as naval reactors and space power and propulsion, as well as research in medical, science, isotope, and electronics arenas. DOE research reactors are aging, and implementing major replacement projects requires long lead times. Primary design drivers include safety, low risk, minimum operation cost, mission flexibility, waste minimization, and long life. Scientists and engineers at the Idaho National Engineering Laboratory are evaluating possible fuel forms, structural materials, reactor geometries, coolants, and moderators

  14. Safety of VVER-440 reactors

    CERN Document Server

    Slugen, Vladimir

    2011-01-01

    Safety of VVER-440 Reactors endeavours to promote an increase in the safety of VVER-440 nuclear reactors via the improvement of fission products limitation systems and the implementation of special non-destructive spectroscopic methods for materials testing. All theoretical and experimental studies performed the by author over the last 25 years have been undertaken with the aim of improving VVER-440 defence in depth, which is one of the most important principle for ensuring safety in nuclear power plants. Safety of VVER-440 Reactors is focused on the barrier system through which the safety pri

  15. Risk-informing safety reviews for non-reactor nuclear facilities: an example application

    International Nuclear Information System (INIS)

    Mubayi, V.; Yue, M.; Bari, R.A.; Azarm, M.A.; Mukaddam, W.; Good, G.; Gonzalez, F.

    2013-01-01

    This paper describes a methodology used to model potential accidents in fuel cycle facilities that employ chemical processes to separate and purify nuclear materials. The methodology is illustrated with an example that uses event and fault trees to estimate the frequency of a specific energetic reaction that can occur in nuclear material processing facilities. The methodology used probabilistic risk assessment (PRA)-related tools as well as information about the chemical reaction characteristics, information on plant design and operational features, and generic data about component failure rates and human error rates. The accident frequency estimates for the specific reaction can be useful to help to risk-inform a safety review process and assess compliance with regulatory requirements. (authors)

  16. Risk-Informing Safety Reviews for Non-Reactor Nuclear Facilities

    International Nuclear Information System (INIS)

    Mubayi, V.; Azarm, A.; Yue, M.; Mukaddam, W.; Good, G.; Gonzalez, F.; Bari, R.A.

    2011-01-01

    This paper describes a methodology used to model potential accidents in fuel cycle facilities that employ chemical processes to separate and purify nuclear materials. The methodology is illustrated with an example that uses event and fault trees to estimate the frequency of a specific energetic reaction that can occur in nuclear material processing facilities. The methodology used probabilistic risk assessment (PRA)-related tools as well as information about the chemical reaction characteristics, information on plant design and operational features, and generic data about component failure rates and human error rates. The accident frequency estimates for the specific reaction help to risk-inform the safety review process and assess compliance with regulatory requirements.

  17. Risk-Informing Safety Reviews for Non-Reactor Nuclear Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Mubayi, V.; Azarm, A.; Yue, M.; Mukaddam, W.; Good, G.; Gonzalez, F.; Bari, R.A.

    2011-03-13

    This paper describes a methodology used to model potential accidents in fuel cycle facilities that employ chemical processes to separate and purify nuclear materials. The methodology is illustrated with an example that uses event and fault trees to estimate the frequency of a specific energetic reaction that can occur in nuclear material processing facilities. The methodology used probabilistic risk assessment (PRA)-related tools as well as information about the chemical reaction characteristics, information on plant design and operational features, and generic data about component failure rates and human error rates. The accident frequency estimates for the specific reaction help to risk-inform the safety review process and assess compliance with regulatory requirements.

  18. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

    This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

  19. Measures ensuring safety of the High Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    1998-04-01

    JAERI has conducted research and development of an HTGR type reactor since 1969 under the project of the multi-purpose high-temperate gas-cooled experimental reactor, whose design was changed to the HTTR in 1985. The reactor license was granted by the Government in 1990 and the construction started next year. Various functions and performances have been tested since 1996 and the initial criticality achieved in 1998. This document consists of six chapters, describing safety matters examined in several development phases. The first chapter deals with succession of the multi-purpose experimental reactor technology and its exchange between JAERI and domestic industries. Chapter 2 reviews new technical findings after the licensing which were reflected to the current safety assessment. These technical items are given in the table form of extensive pages. Chapter 3 and 4 describe the performance tests and the criticality access, respectively. Chapter 5 and 6 deal with the detection of fuel failures and helium gas leaks, respectively. (H.Y.)

  20. Reactor safety method

    International Nuclear Information System (INIS)

    Vachon, L.J.

    1980-01-01

    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

  1. Research on the state-of-the-art of probabilistic safety assessment for non-reactor nuclear facilities (2)

    International Nuclear Information System (INIS)

    Yoshida, Kazuo; Abe, Hitoshi; Yamane, Yuichi; Tashiro, Sinsuke; Muramatsu, Ken

    2007-03-01

    Japan Atomic Energy Agency (JAEA) entrusted with a research on the state-of-the-art of probabilistic safety assessment (PSA) of non-reactor nuclear facilities (NRNF) such as fuel reprocessing and fuel fabrication facilities to the Atomic Energy Society of Japan (AESJ). The objectives of this research is to obtain the basic useful information related for establishing the quantitative performance requirement and for risk-informed regulation through qualifying issues needed to be resolved for applying PSA to NRNF. A special committee of 'Research on the analysis methods for accident consequence in NFRF' was organized by the AESJ. The research activities of the committee were mainly focused on the analysis method for upper bounding consequences of accidents such as events of criticality, explosion, fire and solvent boiling postulated in NRNF resulting in release of radio active material to the environment. This report summarizes the results of research conducted by the committee in FY 2005. (author)

  2. Test results of Run-1 and Run-2 in steam generator safety test facility (SWAT-3)

    International Nuclear Information System (INIS)

    Kurihara, A.; Yatabe, Toshio; Tanabe, Hiromi; Hiroi, Hiroshi

    2003-07-01

    Large leak sodium-water reaction tests were carried out using SWAT-1 rig and SWAT-3 facility in Power Reactor and Nuclear Fuel Development Corporation (PNC) O-arai Engineering Center to obtain the data on the design of the prototype LMFBR Monju steam generator against a large leak accident. This report provides the results of SWAT-3 Runs 1 and 2. In Runs 1 and 2, the heat transfer tube bundle of the evaporator, fabricated by TOSHIBA/IHI, were used, and the pressure relief line was located at the top of evaporator. The water injection rates in the evaporator were 6.7 kg/s and 14.2 (initial)-9.7 kg/s in Runs 1 and 2 respectively, which corresponded to 3.3 tubes and 7.1 (initial)-4.8 tubes failure in actual size system according to iso-velocity modeling. Approximately two hundreds of measurement points were provided to collect data such as pressure, temperature, strain, sodium level, void, thrust load, acceleration, displacement, flow rate, and so on in each run. Initial spike pressures were 1.13 MPa and 2.62 MPa nearest to injection point in Runs 1 and 2 respectively, and the maximum quasi-steady pressures in evaporator were 0.49 MPa and 0.67 MPa in Runs 1 and 2. No secondary tube failure was observed. The rupture disc of evaporator (RD601) burst at 1.1s in Run-1 and at 0.7s in Run-2 after water injected, and the pressure relief system was well-functioned though a few items for improvement were found. (author)

  3. On the safety and performance demonstration tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and validation and verification of computational codes

    International Nuclear Information System (INIS)

    Kim, Jong Bum; Jeong, Ji Young; Lee, Tae Ho; Kim, Sung Kyun; Euh, Dong Jin; Joo, Hyung Kook

    2016-01-01

    The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V and V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V and V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results

  4. On the safety and performance demonstration tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and validation and verification of computational codes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Bum; Jeong, Ji Young; Lee, Tae Ho; Kim, Sung Kyun; Euh, Dong Jin; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V and V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V and V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

  5. FFTF [Fast Flux Test Facility] performance measurements for safety, productivity and control

    International Nuclear Information System (INIS)

    Newland, D.J.; Praetorius, P.R.; Tomaszewski, T.A.

    1987-05-01

    A useful set of performance measurements for Safety, Productivity and Control has evolved at the Fast Flux Test Facility (FFTF). In response to declining budgets and the resulting need to safely manage a manpower rampdown, an ''Early Warning System'' was developed in 1984. Its purpose was to monitor the effects of the staffing rampdown such that appropriate remedial action could be taken to correct adverse trends before a significant problem occurred. 1 tab

  6. Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report October 2014

    Energy Technology Data Exchange (ETDEWEB)

    Ogden, Dan [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-10-01

    Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report October 2014 Highlights • Rory Kennedy, Dan Ogden and Brenden Heidrich traveled to Germantown October 6-7, for a review of the Infrastructure Management mission with Shane Johnson, Mike Worley, Bradley Williams and Alison Hahn from NE-4 and Mary McCune from NE-3. Heidrich briefed the group on the project progress from July to October 2014 as well as the planned path forward for FY15. • Jim Cole gave two invited university seminars at Ohio State University and University of Florida, providing an overview of NSUF including available capabilities and the process for accessing facilities through the peer reviewed proposal process. • Jim Cole and Rory Kennedy co-chaired the NuMat meeting with Todd Allen. The meeting, sponsored by Elsevier publishing, was held in Clearwater, Florida, and is considered one of the premier nuclear fuels and materials conferences. Over 340 delegates attended with 160 oral and over 200 posters presented over 4 days. • Thirty-one pre-applications were submitted for NSUF access through the NE-4 Combined Innovative Nuclear Research Funding Opportunity Announcement. • Fourteen proposals were received for the NSUF Rapid Turnaround Experiment Summer 2014 call. Proposal evaluations are underway. • John Jackson and Rory Kennedy attended the Nuclear Fuels Industry Research meeting. Jackson presented an overview of ongoing NSUF industry research.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-02-01

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

  8. Preliminary Analysis of the Transient Reactor Test Facility (TREAT) with PROTEUS

    Energy Technology Data Exchange (ETDEWEB)

    Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-11-30

    The neutron transport code PROTEUS has been used to perform preliminary simulations of the Transient Reactor Test Facility (TREAT). TREAT is an experimental reactor designed for the testing of nuclear fuels and other materials under transient conditions. It operated from 1959 to 1994, when it was placed on non-operational standby. The restart of TREAT to support the U.S. Department of Energy’s resumption of transient testing is currently underway. Both single assembly and assembly-homogenized full core models have been evaluated. Simulations were performed using a historic set of WIMS-ANL-generated cross-sections as well as a new set of Serpent-generated cross-sections. To support this work, further analyses were also performed using additional codes in order to investigate particular aspects of TREAT modeling. DIF3D and the Monte-Carlo codes MCNP and Serpent were utilized in these studies. MCNP and Serpent were used to evaluate the effect of geometry homogenization on the simulation results and to support code-to-code comparisons. New meshes for the PROTEUS simulations were created using the CUBIT toolkit, with additional meshes generated via conversion of selected DIF3D models to support code-to-code verifications. All current analyses have focused on code-to-code verifications, with additional verification and validation studies planned. The analysis of TREAT with PROTEUS-SN is an ongoing project. This report documents the studies that have been performed thus far, and highlights key challenges to address in future work.

  9. The Integral Test Facility Karlstein

    Directory of Open Access Journals (Sweden)

    Stephan Leyer

    2012-01-01

    Full Text Available The Integral Test Facility Karlstein (INKA test facility was designed and erected to test the performance of the passive safety systems of KERENA, the new AREVA Boiling Water Reactor design. The experimental program included single component/system tests of the Emergency Condenser, the Containment Cooling Condenser and the Passive Core Flooding System. Integral system tests, including also the Passive Pressure Pulse Transmitter, will be performed to simulate transients and Loss of Coolant Accident scenarios at the test facility. The INKA test facility represents the KERENA Containment with a volume scaling of 1 : 24. Component heights and levels are in full scale. The reactor pressure vessel is simulated by the accumulator vessel of the large valve test facility of Karlstein—a vessel with a design pressure of 11 MPa and a storage capacity of 125 m3. The vessel is fed by a benson boiler with a maximum power supply of 22 MW. The INKA multi compartment pressure suppression Containment meets the requirements of modern and existing BWR designs. As a result of the large power supply at the facility, INKA is capable of simulating various accident scenarios, including a full train of passive systems, starting with the initiating event—for example pipe rupture.

  10. Safety of power transformers, power supplies, reactors and similar products - Part 1: General requirements and tests

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1998-01-01

    This International Standard deals with safety aspects of power transformers, power supplies, reactors and similar products such as electrical, thermal and mechanical safety. This standard covers the following types of dry-type transformers, power supplies, including switch mode power supplies, and reactors, the windings of which may be encapsulated or non-encapsulated. It has the status of a group safety publication in accordance with IEC Guide 104.

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

    International Nuclear Information System (INIS)

    Montierth, Leland M.

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-07-19

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

  13. Test and application of thermal neutron radiography facility at Xi'an pulsed reactor

    CERN Document Server

    Yang Jun; Zhao Xiang Feng; Wang Dao Hua

    2002-01-01

    A thermal neutron radiography facility at Xi'an Pulsed Reactor is described as well as its characteristics and application. The experiment results show the inherent unsharpness of BAS ND is 0.15 mm. The efficient thermal neutron n/gamma ratio is lower in not only steady state configuration but also pulsing state configuration and it is improved using Pb filter

  14. Instrumentation of fuel safety test rods of the PWR system in the Phebus reactor

    International Nuclear Information System (INIS)

    Schley, Robert; Leveque, J.P.; Aujollet, J.M.; Dutraive, Pierre; Colome, Jean; Bouly, J.C.

    1979-01-01

    The tests were performed in an experimental cell centred in the core of the PHEBUS water reactor of 50 MW. The CEA make two types of apparatus for testing the safety of PWR fuel. One is for testing a single fuel stick and the other a bunch of 25 sticks. The instrumentation described enables the main parameters of the test to be known: temperatures of the fuel - central temperature of the UO 2 - cladding surface temperatures; temperature of the cooling circuits - thermal balance - temperatures of the structures, etc.; coolant pressure; internal pressure of the fuel sticks; direction and flow rate of the fluid. This instrumentation and the technological problems to be overcome are described and the results of the first tests carried out are given [fr

  15. Nuclear reactor containing facility

    International Nuclear Information System (INIS)

    Hidaka, Masataka; Murase, Michio.

    1994-01-01

    In a reactor containing facility, a condensation means is disposed above the water level of a cooling water pool to condensate steams of the cooling water pool, and return the condensated water to the cooling water pool. Upon occurrence of a pipeline rupture accident, steams generated by after-heat of a reactor core are caused to flow into a bent tube, blown from the exit of the bent tube into a suppression pool and condensated in a suppression pool water, thereby suppressing the pressure in the reactor container. Cooling water in the cooling water pool is boiled by heat conduction due to the condensation of steams, then the steams are exhausted to the outside of the reactor container to remove the heat of the reactor container to the outside of the reactor. In addition, since cooling water is supplied to the cooling water pool quasi-permanently by gravity as a natural force, the reactor container can be cooled by the cooling water pool for a long period of time. Since the condensation means is constituted with a closed loop and interrupted from the outside, radioactive materials are never released to the outside. (N.H.)

  16. Reactor Safety Analysis

    International Nuclear Information System (INIS)

    Arien, B.

    2000-01-01

    The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of two main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents. Main achievements in 1999 are reported

  17. Development of Electrical Capacitance Sensors for Accident Tolerant Fuel (ATF) Testing at the Transient Reactor Test (TREAT) Facility

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Maolong; Ryals, Matthew; Ali, Amir; Blandford, Edward; Jensen, Colby; Condie, Keith; Svoboda, John; O' Brien, Robert

    2016-08-01

    A variety of instruments are being developed and qualified to support the Accident Tolerant Fuels (ATF) program and future transient irradiations at the Transient Reactor Test (TREAT) facility at Idaho National Laboratory (INL). The University of New Mexico (UNM) is working with INL to develop capacitance-based void sensors for determining the timing of critical boiling phenomena in static capsule fuel testing and the volume-averaged void fraction in flow-boiling in-pile water loop fuel testing. The static capsule sensor developed at INL is a plate-type configuration, while UNM is utilizing a ring-type capacitance sensor. Each sensor design has been theoretically and experimentally investigated at INL and UNM. Experiments are being performed at INL in an autoclave to investigate the performance of these sensors under representative Pressurized Water Reactor (PWR) conditions in a static capsule. Experiments have been performed at UNM using air-water two-phase flow to determine the sensitivity and time response of the capacitance sensor under a flow boiling configuration. Initial measurements from the capacitance sensor have demonstrated the validity of the concept to enable real-time measurement of void fraction. The next steps include designing the cabling interface with the flow loop at UNM for Reactivity Initiated Accident (RIA) ATF testing at TREAT and further characterization of the measurement response for each sensor under varying conditions by experiments and modeling.

  18. The advanced test reactor strategic evaluation program

    International Nuclear Information System (INIS)

    Buescher, B.J.

    1989-01-01

    Since the Chernobly accident, the safety of test reactors and irradiation facilities has been critically evaluated from the public's point of view. A systematic evaluation of all safety, environmental, and operational issues must be made in an integrated manner to prioritize actions to maximize benefits while minimizing costs. Such a proactive program has been initiated at the Advanced Test Reactor (ATR). This program, called the Strategic Evaluation Program (STEP), is being conducted for the ATR to provide integrated safety and operational reviews of the reactor against the standards applied to licensed commercial power reactors. This has taken into consideration the lessons learned by the US Nuclear Regulatory Commission (NRC) in its Systematic Evaluation Program (SEP) and the follow-on effort known as the Integrated Safety Assessment Program (ISAP). The SEP was initiated by the NRC to review the designs of older operating nuclear power plants to confirm and document their safety. The ATR STEP objectives are discussed

  19. Research on the state-of-the-art of probabilistic safety assessment for non-reactor nuclear facilities (1)

    International Nuclear Information System (INIS)

    Yoshida, Kazuo; Abe, Hitoshi; Yamane, Yuichi; Tashiro, Sinsuke; Muramatsu, Ken

    2007-02-01

    Japan Atomic Energy Agency (JAEA) entrusted with research on the state-of-the-art of probabilistic safety assessment (PSA) for non-reactor nuclear facilities (NRNF) to the Atomic Energy Society of Japan (AESJ). The objectives of this research is to obtain the basic useful information related for establishing the quantitative performance requirement and for risk-informed regulation through qualifying issues needed to be resolved for applying PSA to NRNF. A special committee of 'research on the analysis methods for accident consequence in NFRF' was organized in the AESJ. The research activities of the committee were mainly focused on the analysis method for upper bounding consequences of accidents such as events of criticality, explosion, fire and solvent boiling postulated in NRNF resulting in release of radio active material to the environment. (author)

  20. Reactor Safety Analysis

    International Nuclear Information System (INIS)

    Arien, B.

    1998-01-01

    The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of four main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents, the development of an expert system for the aid to diagnosis; the development and application of a probabilistic reactor dynamics method. Main achievements in 1999 are reported

  1. Safety Research Experiment Facility Project. Conceptual design report. Volume IV. Reactor containment

    International Nuclear Information System (INIS)

    1975-12-01

    The principal purpose of the SAREF Reactor Containment Building (RCB) is to prevent the uncontrolled release of radioactive materials to the atmosphere as a result of accidental occurrences inside the containment. The RCB houses numerous reactor systems and components including the Prestressed Concrete Reactor Vessel (PCRV). The design of the RCB is of reinforced concrete (steel-lined). The containment building is embedded nearly 100 feet in lava rock. It has therefore been necessary to independently formulate an appropriate and conservative design approach

  2. Fusion technology development: role of fusion facility upgrades and fission test reactors

    International Nuclear Information System (INIS)

    Hsu, P.Y.; Deis, G.A.; Longhurst, G.R.; Miller, L.G.; Schmunk, R.E.

    1983-01-01

    The near term national fusion program is unlikely to follow the aggressive logic of the Fusion Engineering Act of 1980. Faced with level budgets, a large, new fusion facility with an engineering thrust is unlikely in the near future. Within the fusion community the idea of upgrading the existing machines (TFTR, MFTF-B) is being considered to partially mitigate the lack of a design data base to ready the nation to launch an aggressive, mission-oriented fusion program with the goal of power production. This paper examines the cost/benefit issues of using fusion upgrades to develop the technology data base which will be required to support the design and construction of the next generation of fusion machines. The extent of usefulness of the nation's fission test reactors will be examined vis-a-vis the mission of the fusion upgrades. The authors show that while fission neutrons will provide a useful test environment in terms of bulk heating and tritium breeding on a submodule scale, they can play only a supporting role in designing the integrated whole modules and systems to be used in a nuclear fusion machine

  3. Fusion technology development: role of fusion facility upgrades and fission test reactors

    International Nuclear Information System (INIS)

    Hsu, P.Y.; Deis, G.A.; Miller, L.G.; Longhurst, G.R.; Schmunk, R.E.

    1983-01-01

    The near term national fusion program is unlikely to follow the aggressive logic of the Fusion Engineering Act of 1980. Faced with level budgets, a large, new fusion facility with an engineering thrust is unlikely in the near future. Within the fusion community the idea of upgrading the existing machines (TFTR, MFTF-B) is being considered to partially mitigate the lack of a design data base to ready the nation to launch an aggressive, mission-oriented fusion program with the goal of power production. This paper examines the cost/benefit issues of using fusion upgrades to develop the technology data base which will be required to support the design and construction of the next generation of fusion machines. The extent of usefulness of the nation's fission test reactors will be examined vis-a-vis the mission of the fusion upgrades. We will show that while fission neutrons will provide a useful test environment in terms of bulk heating and tritium breeding on a submodule scale, they can play only a supporting role in designing the integrated whole modules and systems to be used in a nuclear fusion machine

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

    International Nuclear Information System (INIS)

    Romano, T.

    1997-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Romano, T.

    1997-09-29

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

  6. Preliminary scoping safety analyses of the limiting design basis protected accidents for the Fast Flux Test Facility tritium production core

    International Nuclear Information System (INIS)

    Heard, F.J.

    1997-01-01

    The SAS4A/SASSYS-l computer code is used to perform a series of analyses for the limiting protected design basis transient events given a representative tritium and medical isotope production core design proposed for the Fast Flux Test Facility. The FFTF tritium and isotope production mission will require a different core loading which features higher enrichment fuel, tritium targets, and medical isotope production assemblies. Changes in several key core parameters, such as the Doppler coefficient and delayed neutron fraction will affect the transient response of the reactor. Both reactivity insertion and reduction of heat removal events were analyzed. The analysis methods and modeling assumptions are described. Results of the analyses and comparison against fuel pin performance criteria are presented to provide quantification that the plant protection system is adequate to maintain the necessary safety margins and assure cladding integrity

  7. Experimental Breeder Reactor II (EBR-II) Fuel-Performance Test Facility (FPTF)

    International Nuclear Information System (INIS)

    Pardini, J.A.; Brubaker, R.C.; Veith, D.J.; Giorgis, G.C.; Walker, D.E.; Seim, O.S.

    1982-01-01

    The Fuel-Performance Test Facility (FPTF) is the latest in a series of special EBR-II instrumented in-core test facilities. A flow control valve in the facility is programmed to vary the coolant flow, and thus the temperature, in an experimental-irradiation subassembly beneath it and coupled to it. In this way, thermal transients can be simulated in that subassembly without changing the temperatures in surrounding subassemblies. The FPTF also monitors sodium flow and temperature, and detects delayed neutrons in the sodium effluent from the experimental-irradiation subassembly beneath it. This facility also has an acoustical detector (high-temperature microphone) for detecting sodium boiling

  8. Evaluation of six papers from France, F.R. Germany, Italy, Japan, UK, USA, on the need for new safety testing facilities to study fuel-failure under loss-of-flow and transient overpower conditions

    International Nuclear Information System (INIS)

    Cicognani, C.; Justin, F.; Jakeman, D.; Hannum, W.H.; Kawaguchi, O.; Kessler, G.

    1975-01-01

    The International Working Group on Fast Reactors invited six countries (France, FRG, Italy, Japan, United Kingdom, USA) to express their views on the importance of fast reactor safety experiments and asked them to give an assessment of the current and future needs of additional safety testing facilities. This papers includes the results of an evaluation of the different views and statements as well as conclusions listed for comparison

  9. Evaluation of six papers from France, F.R. Germany, Italy, Japan, UK, USA, on the need for new safety testing facilities to study fuel-failure under loss-of-flow and transient overpower conditions

    Energy Technology Data Exchange (ETDEWEB)

    Cicognani, C; Justin, F; Jakeman, D; Hannum, W H; Kawaguchi, O; Kessler, G

    1975-08-01

    The International Working Group on Fast Reactors invited six countries (France, FRG, Italy, Japan, United Kingdom, USA) to express their views on the importance of fast reactor safety experiments and asked them to give an assessment of the current and future needs of additional safety testing facilities. This papers includes the results of an evaluation of the different views and statements as well as conclusions listed for comparison.

  10. Establishing a safety and licensing basis for generation IV advanced reactors. License by test

    International Nuclear Information System (INIS)

    Kadak, Andrew C.

    2001-01-01

    The license by test approach to licensing is a novel method of licensing reactors. It provides an opportunity to deal with innovative non-water reactors in a direct way on a time scale that could permit early certification based on tests of a demonstration reactor. The uncertainties in the design and significant contributors to risk would be identified in the PRA during the design. Deterministic analysis computer codes could be tested on a real reactor. Scaling effects and associated uncertainties would be minimized. License by test is an approach that has sufficient merit to be developed and tested

  11. Impact of closed Brayton cycle test results on gas cooled reactor operation and safety

    International Nuclear Information System (INIS)

    Wright, St.A.; Pickard, P.S.

    2007-01-01

    This report summarizes the measurements and model predictions for a series of tests supported by the U.S. Department of Energy that were performed using the recently constructed Sandia Brayton Loop (SBL-30). From the test results we have developed steady-state power operating curves, controls methodologies, and transient data for normal and off-normal behavior, such as loss of load events, and for decay heat removal conditions after shutdown. These tests and models show that because the turbomachinery operates off of the temperature difference (between the heat source and the heat sink), that the turbomachinery can continue to operate (off of sensible heat) for long periods of time without auxiliary power. For our test hardware, operations up to one hour have been observed. This effect can provide significant operations and safety benefits for nuclear reactors that are coupled to a Brayton cycles because the operating turbomachinery continues to provide cooling to the reactor. These capabilities mean that the decay-heat removal can be accommodated by properly managing the electrical power produced by the generator/alternator. In some conditions, it may even be possible to produce sufficient power to continue operating auxiliary systems including the waste heat circulatory system. In addition, the Brayton plant impacts the consequences of off-normal and accident events including loss of load and loss of on-site power. We have observed that for a loss of load or a loss of on-site power event, with a reactor scram, the transient consists initially of a turbomachinery speed increase to a new stable operating point. Because the turbomachinery is still spinning, the reactor is still being cooled provided the ultimate heat sink remains available. These highly desirable operational characteristics were observed in the Sandia Brayton loop. This type of behavior is also predicted by our models. Ultimately, these results provide the designers the opportunity to design gas

  12. An independent safety assessment of Department of Energy nuclear reactor facilities: Training of operating personnel and personnel selection

    International Nuclear Information System (INIS)

    Drain, J.F.

    1981-02-01

    This study has been prepared for the Department of Energy's Nuclear Facilities Personnel Qualification and Training (NFPQT) Committee. Its purpose is to provide the Committee with background information on, and assessment of, the selection, training, and qualification of nuclear reactor operating personnel at DOE-owned facilities

  13. Radiation safety aspects during nondestructive testing of reactor shielding components by gamma radiometry

    International Nuclear Information System (INIS)

    Viswanathan, S.; Jose, M.T.; Venkatraman, B.

    2016-01-01

    In nuclear facilities, effective shielding of radioactive components and structures are essential to ensure radiation protection to operating personnel. The shield structures are made of lead, steel and concrete with varying thickness of up to 1200 mm. It needs to be verified for shielding integrity, presence of voids, blowholes and defects to avoid exposure to workers and to public at large. Radiometry using gamma source serves as excellent tool for non-destructive examination of such structures and components. Gamma sources of high activity up to 50 Curies (gamma camera type) depending on the thickness of component have to be used. During the testing exposure to the operating personnel needs to be minimized, this requires certain safety procedures to be followed. This paper focuses the methodology to be adapted by means of selection of source, effective training of personnel, compliance with safety requirements and maintenance of source devices

  14. Use of EBR-II as a principal fast breeder reactor irradiation test facility in the U.S

    International Nuclear Information System (INIS)

    Staker, R.G.; Seim, O.S.; Beck, W.N.; Golden, G.H.; Walters, L.C.

    1975-01-01

    The EBR-II as originally designed and operated by the Argonne National Laboratory was successful in demonstrating the operation of a sodium-cooled fast breeder power plant with a closed fuel reprocessing cycle. Subsequent operation has been as an experimental facility where thousands of irradiation tests have been performed. Conversion to this application entailed the design and fabrication of special irradiation subassemblies for in-core irradiations, additions to existing facilities for out-of-core irradiations, and additions to existing facilities for out-of-core experiments. Experimental subassemblies now constitute about one third of the core, and changes in the core configuration occur about monthly, requiring neutronic and thermal-hydraulics analyses and monitoring of the reactor dynamic behavior. The surveillance programs provided a wealth of information on irradiation induced swelling and creep, in-reactor fracture behavior, and the compatibility of materials with liquid sodium. (U.S.)

  15. Initial characterization of the ATR [Advanced Test Reactor] Large Gamma Facility

    International Nuclear Information System (INIS)

    Schnitzler, B.G.; Rogers, J.W.

    1986-05-01

    Radiation fields in the ATR Large Gamma Facility test volume are characterized. The preliminary characterization efforts described in this report include total dose rate measurements in the facility, development of a simple methodology for calculating radiation fields from the ATR fuel element power histories, and a comparison of the measured and calculated values

  16. Fire protection at the Fast Flux Test Facility (a sodium cooled test reactor)

    International Nuclear Information System (INIS)

    Bell, J.R.

    1980-01-01

    For purposes of this presentation, fire protection at the FFTF is subdivided into two catagories; protection for non-sodium areas and protection for areas containing sodium. Fire protection systems and philosophies for non-sodium areas at the FFTF are very similar to those used at conventional power plants being constructed throughout the country. They follow, essentially, the NRC rules and guidelines and ANSI 59.4 Generic Requirements for Light Water Nuclear Power Plant Fire Protection. The FFTF with its support facilities have their own water system comprised of a looped 8'' and 10'' underground distribution system, three 1500 GPM fire pumps and three ground level storage tanks totaling 736,000 gallons with 420,000 reserved for fire protection. Fire hydrants are enclosed with hose houses outfitted for use by the Emergency Response Team (ERT). Fire prevention systems for sodium areas of the FFTF are also described

  17. Probabilistic safety assessment for food irradiation facility

    International Nuclear Information System (INIS)

    Solanki, R.B.; Prasad, M.; Sonawane, A.U.; Gupta, S.K.

    2012-01-01

    Highlights: ► Different considerations are required in PSA for Non-Reactor Nuclear Facilities. ► We carried out PSA for food irradiation facility as a part of safety evaluation. ► The results indicate that the fatal exposure risk is below the ‘acceptable risk’. ► Adequate operator training and observing good safety culture would reduce the risk. - Abstract: Probabilistic safety assessment (PSA) is widely used for safety evaluation of Nuclear Power Plants (NPPs) worldwide. The approaches and methodologies are matured and general consensus exists on using these approaches in PSA applications. However, PSA applications for safety evaluation for non-reactor facilities are limited. Due to differences in the processes in nuclear reactor facilities and non-reactor facilities, the considerations are different in application of PSA to these facilities. The food irradiation facilities utilize gamma irradiation sources, X-ray machines and electron accelerators for the purpose of radiation processing of variety of food items. This is categorized as Non-Reactor Nuclear Facility. In this paper, the application of PSA to safety evaluation of food irradiation facility is presented considering the ‘fatality due to radiation overexposure’ as a risk measure. The results indicate that the frequency of the fatal exposure is below the numerical acceptance guidance for the risk to the individual. Further, it is found that the overall risk to the over exposure can be reduced by providing the adequate operator training and observing good safety culture.

  18. Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report November 2014

    Energy Technology Data Exchange (ETDEWEB)

    Soelberg, Renae [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-11-01

    Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report November 2014 Highlights Rory Kennedy and Sarah Robertson attended the American Nuclear Society Winter Meeting and Nuclear Technology Expo in Anaheim, California, Nov. 10-13. ATR NSUF exhibited at the technology expo where hundreds of meeting participants had an opportunity to learn more about ATR NSUF. Dr. Kennedy briefed the Nuclear Engineering Department Heads Organization (NEDHO) on the workings of the ATR NSUF. • Rory Kennedy, James Cole and Dan Ogden participated in a reactor instrumentation discussion with Jean-Francois Villard and Christopher Destouches of CEA and several members of the INL staff. • ATR NSUF received approval from the NE-20 office to start planning the annual Users Meeting. The meeting will be held at INL, June 22-25. • Mike Worley, director of the Office of Innovative Nuclear Research (NE-42), visited INL Nov. 4-5. Milestones Completed • Recommendations for the Summer Rapid Turnaround Experiment awards were submitted to DOE-HQ Nov. 12 (Level 2 milestone due Nov. 30). Major Accomplishments/Activities • The University of California, Santa Barbara 2 experiment was unloaded from the GE-2000 at HFEF. The experiment specimen packs will be removed and shipped to ORNL for PIE. • The Terrani experiment, one of three FY 2014 new awards, was completed utilizing the Advanced Photon Source MRCAT beamline. The experiment investigated the chemical state of Ag and Pd in SiC shell of irradiated TRISO particles via X-ray Absorption Fine Structure (XAFS) spectroscopy. Upcoming Meetings/Events • The ATR NSUF program review meeting will be held Dec. 9-10 at L’Enfant Plaza. In addition to NSUF staff and users, NE-4, NE-5 and NE-7 representatives will attend the meeting. Awarded Research Projects Boise State University Rapid Turnaround Experiments (14-485 and 14-486) Nanoindentation and TEM work on the T91, HT9, HCM12A and 9Cr ODS specimens has been completed at

  19. Safety assessment document for spent fuel handling, packaging, and storage demonstrations at the E-MAD facility on the Nevada Test Site

    International Nuclear Information System (INIS)

    1985-04-01

    The objectives for spent fuel handling and packaging demonstration are to develop the capability to satisfactorily encapsulate typical commercial nuclear reactor spent fuel assemblies and to establish the suitability of interim dry surface and near surface storage concepts. To accomplish these objectives, spent fuel assemblies from a pressurized water reactor have been received, encapsulated in steel canisters, and emplaced in on-site storage facilities and subjected to other tests. As an essential element of these demonstrations, a thorough safety assessment of the demonstration activities conducted at the E-MAD facility has been completed. This document describes the site location and characteristics, the existing E-MAD facility, and the facility modifications and equipment additions made specifically for the demonstrations. The document also summarizes the Quality Assurance Program utilized, and specifies the principal design criteria applicable to the facility modifications, equipment additions, and process operations. Evaluations have been made of the radiological impacts of normal operations, abnormal operations, and postulated accidents. Analyses have been performed to determine the affects on nuclear criticality safety of postulated accidents and credible natural phenomena. The consequences of postulated accidents resulting in fission product gas release have also been estimated. This document identifies the engineered safety features, procedures, and site characteristics that (1) prevent the occurrence of potential accidents or (2) assure that the consequences of postulated accidents are either insignificant or adequately mitigated

  20. Reactor system safety assurance

    International Nuclear Information System (INIS)

    Mattson, R.J.

    1984-01-01

    The philosophy of reactor safety is that design should follow established and conservative engineering practices, there should be safety margins in all modes of plant operation, special systems should be provided for accidents, and safety systems should have redundant components. This philosophy provides ''defense in depth.'' Additionally, the safety of nuclear power plants relies on ''safety systems'' to assure acceptable response to design basis events. Operating experience has shown the need to study plant response to more frequent upset conditions and to account for the influence of operators and non-safety systems on overall performance. Defense in depth is being supplemented by risk and reliability assessment

  1. Nuclear reactor facility

    International Nuclear Information System (INIS)

    Wampole, N.C.

    1978-01-01

    In order to improve the performance of manitenance and inspections it is proposed for a nuclear reactor facility with a primary circuit containing liquid metal to provide a thermally insulated chamber, within which are placed a number of components of the primary circuit, as e.g. valves, recirculation pump, heat exchangers. The isolated placement permit controlled preheating on one hand, but prevents undesirable heating of adjacent load-bearing elements on the other. The chamber is provided with heating devices and, on the outside, with cooling devices; it is of advantage to fill it with an inert gas. (UWI) 891 HP [de

  2. Quality assurance: an independent safety assessment of Department of Energy nuclear reactor facilities

    International Nuclear Information System (INIS)

    Frankhouser, W.L.; Bass, W. Jr.; Langston, M.E.

    1981-02-01

    This report presents the assessments of QA programs at eight DOE reactor sites as performed by three members of the NFPQT Support Team. A summation of assessments is presented. That summation includes discussion of findings and recommendations for follow-on actions. The detailed record of contractor-by-contractor reviews is provided. A discussion of the approach in performing the QA assessments is presented, and the basis and limitations of the assessments are discussed

  3. Accidents and failures in reactor facilities for test and research and reactor facilities in the stage of research and development in fiscal year 1987

    International Nuclear Information System (INIS)

    1988-01-01

    The number of accidents and failures reported in fiscal year 1987 in conformity with the law on the regulation of nuclear reactors and others was three. One case occurred during operation, and two cases occurred in shutdown state. One case was caused by improper construction management, and two cases were due to improper maintenance management. The effect of radioactivity to the surrounding environment of reactor facilities due to these accidents and failures did not arise. These occurred in the NSRR of Japan Atomic Energy Research Institute (Tokai), the experimental FBR Joyo and the ATR Fugen Power Station of Power Reactor and Nuclear Fuel Development Corp. In addition to these, the light troubles reported on the basis of the notice from the director of Science and Technology Agency dated September 1, 1981, were three cases. (K.I.)

  4. Reactor Safety Assessment System

    International Nuclear Information System (INIS)

    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). RSAS 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

  5. Reactor safety assessment system

    International Nuclear Information System (INIS)

    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

  6. Structural safety of HDR reactor building during large scale vibration tests

    International Nuclear Information System (INIS)

    Stangenberg, F.; Zinn, R.

    1985-01-01

    In the second phase of the HDR investigations, a high shaker excitation of the building is planned using a large shaker which will be located on the operating floor and will be brought up to speed in a balanced condition and then unbalanced and decoupled from the drive system. With decreasing speed the shaker comes in resonance with the building frequencies and its energy is transferred to the building. In this paper the structural safety of the reactor building during the projected shaker tests is analysed. Dynamic response calculations with coupling between building and shaker by simultaneously integrating the equilibrium equations of both building and shaker are presented. The resulting building stresses, soil pressures etc. are compared with allowable values. (orig.)

  7. Safety Evaluation for Packaging for onsite Transfer of plutonium recycle test reactor ion exchange columns

    International Nuclear Information System (INIS)

    Smith, R.J.

    1995-01-01

    The purpose of this Safety Evaluation for Packaging (SEP) is to authorize the use of three U.S. Department of Transportation (DOT) 7A, Type A metal boxes (Capital Industries Part No. S 0600-0600-1080- 0104) to package 12 Plutonium Recycle Test Reactor (PRTR) ion exchange columns as low-level waste (LLW). The packages will be transferred from the 309 Building in the 300 Area to low level waste burial in the 200 West Area. Revision 1 of WHC-SD-TP-SEP-035 (per ECN No. 621467) documents that the boxes containing ion exchange columns and grout will maintain the payload under normal conditions of transport if transferred without the box lids

  8. Verification of reactor safety codes

    International Nuclear Information System (INIS)

    Murley, T.E.

    1978-01-01

    The safety evaluation of nuclear power plants requires the investigation of wide range of potential accidents that could be postulated to occur. Many of these accidents deal with phenomena that are outside the range of normal engineering experience. Because of the expense and difficulty of full scale tests covering the complete range of accident conditions, it is necessary to rely on complex computer codes to assess these accidents. The central role that computer codes play in safety analyses requires that the codes be verified, or tested, by comparing the code predictions with a wide range of experimental data chosen to span the physical phenomena expected under potential accident conditions. This paper discusses the plans of the Nuclear Regulatory Commission for verifying the reactor safety codes being developed by NRC to assess the safety of light water reactors and fast breeder reactors. (author)

  9. Safety Management at PUSPATI TRIGA Reactor (RTP)

    International Nuclear Information System (INIS)

    Ligam, A.S.; Zarina Masood; Ahmad Nabil Abdul Rahim

    2011-01-01

    Adequate safety measures and precautions, which follow relevant safety standards and procedures, should be in place so that personnel safety is assured. Nevertheless, the public, visitor, contractor or anyone who wishes to enter or be in the reactor building should be well informed with the safety measures applied. Furthermore, these same elements of safety are also applied to other irradiation facilities within the premises of Nuclear Malaysia. This paper will describes and explains current safety management system being enforced especially in the TRIGA PUSPATI Reactor (RTP) namely radiation monitoring system, safety equipment, safe work instruction, and interconnected internal and external health, safety and security related departments. (author)

  10. Nuclear reactor safety systems

    International Nuclear Information System (INIS)

    Ball, R.M.; Roberts, R.C.

    1980-01-01

    A safety system for shutting down a nuclear reactor under overload conditions is described. The system includes a series of parallel-connected computer memory type look-up tables each of which receives data on a particular reactor parameter and in each of which a precalculated functional value for that parameter is stored indicative of the percentage of maximum reactor load that the parameter contributes. The various functional values corresponding to the actual measured parameters are added together to provide a control signal used to shut down the reactor under overload conditions. (U.K.)

  11. Large-scale experimental facility for emergency condition investigation of a new generation NPP WWER-640 reactor with passive safety systems

    International Nuclear Information System (INIS)

    Aniskevich, Y.N.; Vasilenko, V.A.; Zasukha, V.K.; Migrov, Y.A.; Khabensky, V.B.

    1997-01-01

    The creation of the large-scale integral experimental facility (KMS) is specified by the programme of the experimental investigations to justify the engineering decisions on the safety of the design of the new generation NPP with the reactor WWER-640. The construction of KMS in a full volume will allow to conduct experimental investigations of all physical phenomena and processes, practically, occurring during the accidents on the NPPs with the reactor of WWER type and including the heat - mass exchange processes with low rates of the coolant, which is typical during the utilization of the passive safety systems, process during the accidents with a large leak, and also the complex intercommunicated processes in the reactor unit, passive safety systems and in the containment with the condition of long-term heat removal to the final absorber. KMS is being constructed at the Research Institute of Technology (NITI), Sosnovy Bor, Leningrad region, Russia. (orig.)

  12. Large-scale experimental facility for emergency condition investigation of a new generation NPP WWER-640 reactor with passive safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Aniskevich, Y.N.; Vasilenko, V.A.; Zasukha, V.K.; Migrov, Y.A.; Khabensky, V.B. [Research Inst. of Technology NITI (Russian Federation)

    1997-12-31

    The creation of the large-scale integral experimental facility (KMS) is specified by the programme of the experimental investigations to justify the engineering decisions on the safety of the design of the new generation NPP with the reactor WWER-640. The construction of KMS in a full volume will allow to conduct experimental investigations of all physical phenomena and processes, practically, occurring during the accidents on the NPPs with the reactor of WWER type and including the heat - mass exchange processes with low rates of the coolant, which is typical during the utilization of the passive safety systems, process during the accidents with a large leak, and also the complex intercommunicated processes in the reactor unit, passive safety systems and in the containment with the condition of long-term heat removal to the final absorber. KMS is being constructed at the Research Institute of Technology (NITI), Sosnovy Bor, Leningrad region, Russia. (orig.). 5 refs.

  13. Large-scale experimental facility for emergency condition investigation of a new generation NPP WWER-640 reactor with passive safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Aniskevich, Y N; Vasilenko, V A; Zasukha, V K; Migrov, Y A; Khabensky, V B [Research Inst. of Technology NITI (Russian Federation)

    1998-12-31

    The creation of the large-scale integral experimental facility (KMS) is specified by the programme of the experimental investigations to justify the engineering decisions on the safety of the design of the new generation NPP with the reactor WWER-640. The construction of KMS in a full volume will allow to conduct experimental investigations of all physical phenomena and processes, practically, occurring during the accidents on the NPPs with the reactor of WWER type and including the heat - mass exchange processes with low rates of the coolant, which is typical during the utilization of the passive safety systems, process during the accidents with a large leak, and also the complex intercommunicated processes in the reactor unit, passive safety systems and in the containment with the condition of long-term heat removal to the final absorber. KMS is being constructed at the Research Institute of Technology (NITI), Sosnovy Bor, Leningrad region, Russia. (orig.). 5 refs.

  14. Nuclear power reactor safety

    International Nuclear Information System (INIS)

    Pon, G.A.

    1976-10-01

    This report is based on the Atomic Energy of Canada Limited submission to the Royal Commission on Electric Power Planning on the safety of CANDU reactors. It discusses normal operating conditions, postulated accident conditions, and safety systems. The release of radioactivity under normal and accident conditions is compared to the limits set by the Atomic Energy Control Regulations. (author)

  15. SRP reactor safety evolution

    International Nuclear Information System (INIS)

    Rankin, D.B.

    1984-01-01

    The Savannah River Plant reactors have operated for over 100 reactor years without an incident of significant consequence to on or off-site personnel. The reactor safety posture incorporates a conservative, failure-tolerant design; extensive administrative controls carried out through detailed operating and emergency written procedures; and multiple engineered safety systems backed by comprehensive safety analyses, adapting through the years as operating experience, changes in reactor operational modes, equipment modernization, and experience in the nuclear power industry suggested. Independent technical reviews and audits as well as a strong organizational structure also contribute to the defense-in-depth safety posture. A complete review of safety history would discuss all of the above contributors and the interplay of roles. This report, however, is limited to evolution of the engineered safety features and some of the supporting analyses. The discussion of safety history is divided into finite periods of operating history for preservation of historical perspective and ease of understanding by the reader. Programs in progress are also included. The accident at Three Mile Island was assessed for its safety implications to SRP operation. Resulting recommendations and their current status are discussed separately at the end of the report. 16 refs., 3 figs

  16. Multimegawatt Space Reactor Safety

    International Nuclear Information System (INIS)

    Stanley, M.L.

    1989-01-01

    The Multimegawatt (MMW) Space Reactor Project supports the Strategic Defense Initiative Office requirement to provide reliable, safe, cost-effective, electrical power in the MMW range. Specifically, power may be used for neutral particle beams, free electron lasers, electromagnetic launchers, and orbital transfer vehicles. This power plant technology may also apply to the electrical power required for other uses such as deep-space probes and planetary exploration. The Multimegawatt Space Reactor Project, the Thermionic Fuel Element Verification Program, and Centaurus Program all support the Multimegawatt Space Nuclear Power Program and form an important part of the US Department of Energy's (DOE's) space and defense power systems activities. A major objective of the MMW project is the development of a reference flight system design that provides the desired levels of public safety, health protection, and special nuclear material (SNM) protection when used during its designated missions. The safety requirements for the MMW project are a hierarchy of requirements that consist of safety requirements/regulations, a safety policy, general safety criteria, safety technical specifications, safety design specifications, and the system design. This paper describes the strategy and philosophy behind the development of the safety requirements imposed upon the MMW concept developers. The safety organization, safety policy, generic safety issues, general safety criteria, and the safety technical specifications are discussed

  17. Safety Evaluation Report related to the renewal of the operating license for the General Electric-Nuclear Test Reactor (GE-NTR) (Docket No. 50-73)

    International Nuclear Information System (INIS)

    1984-09-01

    This Safety Evaluation Report for the application filed by the General Electric Company (GE) for a renewal license number R-33 to continue to operate its research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by GE and is located in Pleasanton, California. The staff concludes that the reactor can continue to be operated by GE without endangering the health and safety of the public

  18. Development of Safety Review Guidance for Research and Training Reactors

    International Nuclear Information System (INIS)

    Oh, Kju-Myeng; Shin, Dae-Soo; Ahn, Sang-Kyu; Lee, Hoon-Joo

    2007-01-01

    The KINS already issued the safety review guidance for pressurized LWRs. But the safety review guidance for research and training reactors were not developed. So, the technical standard including safety review guidance for domestic research and training reactors has been applied mutates mutandis to those of nuclear power plants. It is often difficult for the staff to effectively perform the safety review of applications for the permit by the licensee, based on peculiar safety review guidance. The NRC and NSC provide the safety review guidance for test and research reactors and European countries refer to IAEA safety requirements and guides. The safety review guide (SRG) of research and training reactors was developed considering descriptions of the NUREG- 1537 Part 2, previous experiences of safety review and domestic regulations for related facilities. This study provided the safety review guidance for research and training reactors and surveyed the difference of major acceptance criteria or characteristics between the SRG of pressurized light water reactor and research and training reactors

  19. Reactor cold neutron source facility, the first in Japan

    International Nuclear Information System (INIS)

    Utsuro, Masahiko; Maeda, Yutaka; Kawai, Takeshi; Tashiro, Tameyoshi; Sakakibara, Shoji; Katada, Minoru.

    1986-01-01

    In the Research Reactor Institute, Kyoto University, the first cold neutron source facility for the reactor in Japan was installed, and various tests are carried out outside the reactor. Nippon Sanso K.K. had manufactured it. After the prescribed tests outside the reactor, this facility will be installed soon in the reactor, and its outline is described on this occasion. Cold neutrons are those having very small energy by being cooled to about-250 deg C. Since the wavelength of the material waves of cold neutrons is long, and their energy is small, they are very advantageous as an experimental means for clarifying the structure of living body molecules and polymers, the atom configuration in alloys, and atomic and molecular movements by neutron scattering and neutron diffraction. The basic principle of the cold neutron source facility is to irradiate thermal neutrons on a cold moderator kept around 20 K, and to moderate and cool the neutrons by nuclear scattering to convert to cold neutrons. The preparatory research on cold neutrons and hydrogen liquefaction, the basic design to put the cold neutron source facility in the graphite moderator facility, the safety countermeasures, the manufacture and quality control, the operation outside the reactor and the performance are reported. The cold neutron source facility comprises a cold moderator tank and other main parts, a deuterium gas tank, a helium refrigerator and instrumentation. (Kako, I.)

  20. On the Safety and Performance Demonstration Tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and Validation and Verification of Computational Codes

    Directory of Open Access Journals (Sweden)

    Jong-Bum Kim

    2016-10-01

    Full Text Available The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR has been developed and the validation and verification (V&V activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1, produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

  1. An experimental test facility to support development of the fluoride-salt-cooled high-temperature reactor

    International Nuclear Information System (INIS)

    Yoder, Graydon L.; Aaron, Adam; Cunningham, Burns; Fugate, David; Holcomb, David; Kisner, Roger; Peretz, Fred; Robb, Kevin; Wilgen, John; Wilson, Dane

    2014-01-01

    Highlights: • • A forced convection test loop using FLiNaK salt was constructed to support development of the FHR. • The loop is built of alloy 600, and operating conditions are prototypic of expected FHR operation. • The initial test article is designed to study pebble bed heat transfer cooled by FLiNaK salt. • The test facility includes silicon carbide test components as salt boundaries. • Salt testing with silicon carbide and alloy 600 confirmed acceptable loop component lifetime. - Abstract: The need for high-temperature (greater than 600 °C) energy transport systems is significantly increasing as the world strives to improve energy efficiency and develop alternatives to petroleum-based fuels. Liquid fluoride salts are one of the few energy transport fluids that have the capability of operating at high temperatures in combination with low system pressures. The fluoride-salt-cooled high-temperature reactor design uses fluoride salt to remove core heat and interface with a power conversion system. Although a significant amount of experimentation has been performed with these salts, specific aspects of this reactor concept will require experimental confirmation during the development process. The experimental facility described here has been constructed to support the development of the fluoride-salt-cooled high-temperature reactor concept. The facility is capable of operating at up to 700 °C and incorporates a centrifugal pump to circulate FLiNaK salt through a removable test section. A unique inductive heating technique is used to apply heat to the test section, allowing heat transfer testing to be performed. An air-cooled heat exchanger removes added heat. Supporting loop infrastructure includes a pressure control system, a trace heating system, and a complement of instrumentation to measure salt flow, temperatures, and pressures around the loop. The initial experiment is aimed at measuring fluoride-salt heat transfer inside a heated pebble bed

  2. Reactor safety analysis

    International Nuclear Information System (INIS)

    Arien, B.

    1998-01-01

    Risk assessments of nuclear installations require accurate safety and reliability analyses to estimate the consequences of accidental events and their probability of occurrence. The objective of the work performed in this field at the Belgian Nuclear Research Centre SCK-CEN is to develop expertise in probabilistic and deterministic reactor safety analysis. The four main activities of the research project on reactor safety analysis are: (1) the development of software for the reliable analysis of large systems; (2) the development of an expert system for the aid to diagnosis; (3) the development and the application of a probabilistic reactor-dynamics method, and (4) to participate in the international PHEBUS-FP programme for severe accidents. Progress in research during 1997 is described

  3. Safety aspects of reprocessing and plutonium fuel facilities in power reactor and nuclear fuel development corporation

    International Nuclear Information System (INIS)

    Sato, S.; Akutsu, H.; Nakajima, K.; Kono, K.; Muto, T.

    1977-01-01

    PNC completed the construction of the first Japanese reprocessing plant in 1974, and the startup is now under way. The plant will have a capacity of 0.7 metric tons of spent fuel per day. Various safety measures for earthquake, radiation, criticality, fire, explosion and leakage of radioactive materials are provided in the plant. 8,000 Ci of Kr-85 and 50 Ci of H-3 per day will be released from the plant to enviroment. Skin dose is conservatively estimated to be about 30 mrem per year. Liquid waste containing 0.7 Ci per day will be discharged into the sea. Whole body dose is conservatively estimated to be 10 mrem per year. R and D for removal of Kr-85 and reducing radioactivity released into the sea are being carried out. Developmental works for solidification of radioactive liquid waste are also being conducted. Safety control in plutonium handling work for both R and D and fuel fabrication has been successfully conducted without significant abnormal occurrence in these ten years. By ''zero-contamination control policy'', surface contamination and airborne contamination in operation rooms are maintained at the background level in usual operation. The intake of plutonium was found at the maximum about one-hundredths of the MPB. External exposure has been generally controlled below three-tenths rem for three months, by shielding and mechanization of process. The radioactivity concentration of exhaust air and liquid effluent disposal is ensured far below the regulation level. Nuclear material control is maintained by a computer system, and no criticality problem has occurred. The safeguard system and installation has been improved, and is sufficient to satisfy the IAEA regulation

  4. Development of the test facilities for the measurement of core flow and pressure distribution of SMART reactor

    International Nuclear Information System (INIS)

    Ko, Y.J.; Euh, D.J.; Youn, Y.J.; Chu, I.C.; Kwon, T.S.

    2011-01-01

    A design of SMART reactor has been developed, of which the primary system is composed of four internal circulation pumps, a core of 57 fuel assemblies, eight cassettes of steam generators, flow mixing head assemblies, and other internal structures. Since primary design features are very different from conventional reactors, the characteristics of flow and pressure distribution are expected to be different accordingly. In order to analyze the thermal margin and hydraulic design characteristics of SMART reactor, design quantification tests for flow and pressure distribution with a preservation of flow geometry are necessary. In the present study, the design feature of the test facility in order to investigate flow and pressure distribution, named “SCOP” is described. In order to preserve the flow distribution characteristics, the SCOP is linearly reduced with a scaling ratio of 1/5. The core flow rate of each fuel assembly is measured by a venturi meter attached in the lower part of the core simulator having a similarity of pressure drop for nominally scaled flow conditions. All the 57 core simulators and 8 S/G simulators are precisely calibrated in advance of assembling in test facilities. The major parameters in tests are pressures, differential pressures, and core flow distribution. (author)

  5. Nuclear reactor safety system

    International Nuclear Information System (INIS)

    Ball, R.M.; Roberts, R.C.

    1983-01-01

    The invention provides a safety system for a nuclear reactor which uses a parallel combination of computer type look-up tables each of which receives data on a particular parameter (from transducers located in the reactor system) and each of which produces the functional counterpart of that particular parameter. The various functional counterparts are then added together to form a control signal for shutting down the reactor. The functional counterparts are developed by analysis of experimental thermal and hydraulic data, which are used to form expressions that define safe conditions

  6. Inquiry relating to safety due to modification of usage of nuclear fuel material (establishment of waste safety testing facility) in Tokai Laboratory, Japan Atomic Energy Research Institute

    International Nuclear Information System (INIS)

    1979-01-01

    Application was made to the director of the Science and Technology Agency (STA) for the license relating to the modification of usage of nuclear fuel material (the establishment of waste safety testing facility) from the director of the Japan Atomic Energy Research Institute on November 30, 1978. After passing through the safety evaluation in the Nuclear Safety Bureau of STA, inquiry was conducted to the head of the Atomic Energy Safety Commission (AESC) on June 6, 1979, from the director of the STA. The head of AESC directed to conduct the safety examination to the head of the Nuclear Fuel Safety Examination Specialist Committee on June 7, 1979. The content of the modification of usage of nuclear fuel material is the establishment of waste safety testing facility to study and test the safety relating to the treatment and disposal of high level radioactive liquid wastes due to the reprocessing of spent fuel. As for the results of the safety examination, the siting of the waste safety testing facility which is located in the Tokai Laboratory, Japan Atomic Energy Research Institute (JAERI), and the test plan of the glass solidification of high level radioactive liquid are presented as the outline of the study plan. The building, main equipments including six cells, the isolation room and the glove box, the storage, and the disposal facilities for gas, liquid and solid wastes are explained as the outline of the facilities. Concerning the items from the viewpoint of safety, aseismatic design, slightly vacuum operation, shielding, decay heat removal, fire protection, explosion protection, criticality management, radiation management and environmental effect were evaluated, and the safety was confirmed. (Nakai, Y.)

  7. Space nuclear reactor safety

    International Nuclear Information System (INIS)

    Damon, D.; Temme, M.; Brown, N.

    1990-01-01

    Definition of safety requirements and design features of the SP-100 space reactor power system has been guided by a mission risk analysis. The analysis quantifies risk from accidental radiological consequences for a reference mission. Results show that the radiological risk from a space reactor can be made very low. The total mission risk from radiological consequences for a shuttle-launched, earth orbit SP-100 mission is estimated to be 0.05 Person-REM (expected values) based on a 1 mREM/yr de Minimus dose. Results are given for each mission phase. The safety benefits of specific design features are evaluated through risk sensitivity analyses

  8. Comparison of in-plant performance test data with analytic prediction of reactor safety system injection transient (U)

    International Nuclear Information System (INIS)

    Roy, B.N.; Neill, C.H. Jr.

    1993-01-01

    This paper compares the performance test data from injection transients for both of the subsystems of the Supplementary Safety System of the Savannah River Site production reactor with analytical predictions from an in-house thermal hydraulic computer code. The code was initially developed for design validation of the new Supplementary Safety System subsystem, but is shown to be equally capable of predicting the performance of the Supplementary Safety System existing subsystem even though the two subsystem transient injections have marked differences. The code itself was discussed and its validation using prototypic tests with simulated fluids was reported in an earlier paper (Roy and Nomm 1991)

  9. The engineering test facility

    International Nuclear Information System (INIS)

    Steiner, D.; Becraft, W.R.; Sager, P.H.

    1981-01-01

    The vehicle by which the fusion program would move into the engineering testing phase of fusion power development is designated the Engineering Test Facility (ETF). The ETF would provide a test-bed for reactor components in the fusion environment. In order to initiate preliminary planning for the ETF decision, the Office of Fusion Energy established the ETF Design Center activity to prepare the design of the ETF. This paper describes the design status of the ETF. (orig.)

  10. Engineering test facility

    International Nuclear Information System (INIS)

    Steiner, D.; Becraft, W.R.; Sager, P.H.

    1981-01-01

    The vehicle by which the fusion program would move into the engineering testing phase of fusion power development is designated the Engineering Test Facility (ETF). The ETF would provide a test-bed for reactor components in the fusion environment. In order to initiate preliminary planning for the ETF decision, the Office of Fusion Energy established the ETF Design Center activity to prepare the design of the ETF. This paper described the design status of the ETF

  11. Reactor feedwater facility

    Energy Technology Data Exchange (ETDEWEB)

    Fujii, Tadashi; Kinoshita, Shoichiro; Akatsu, Jun-ichi

    1996-04-30

    In a reactor feedwater facility in which one stand-by system and at least three ordinary systems are disposed in parallel, each of the feedwater pumps is driven by an electromotor, and has substantially the same capacity. At least two systems among the ordinary systems have a pump rotation number variable means. Since the volume of each of the feedwater pump of each system is determined substantially equal, standardization is enabled to facilitate the production. While the number of electromotors is increased, since they are driven by electromotors, turbines, steam pipelines and valves for driving feed water pumps can be eliminated. Therefore, the feedwater pumps can be disposed to a region of low radiation dose being separated from a main turbine and a main condensator, to improve the degree of freedom in view of the installation. In addition, accessibility to equipments during operation is improved to improve the maintenance of feed water facilities. The number of parts for equipments can be reduced compared with that in a turbine-driving system thereby capable of reducing the operation amount for the maintenance and inspection. (N.H.)

  12. Department of Energy's Advanced Test Reactor (ATR), July 14--18, 1980: An independent on-site safety review

    International Nuclear Information System (INIS)

    1981-02-01

    The intent of this review was not to conduct a detailed in-depth audit, but rather to make a broad management assessment of ATR operations. The results of the review should only be considered as having identified trends or indications. The Team's observations and recommendations for the most part are based upon standards used for licensed reactor facility practices. These standards form the basis for many of the comments in this report. The Team believes that a uniform minimum standard of performance should be achieved in the operation of DOE reactors. In order to assure that this is accomplished, clear standards are necessary. Consistent with the past AEC and ERDA policy, the Team has used the standards of the commercial nuclear power industry. It is recognized that this approach is conservative, in that the ATR reactor has a significantly greater degree of inherent safety (lower pressure, temperature, power, etc.) than a licensed reactor. Although the Review Team found no indications or evidence that the plant is being operated in an unsafe manner, various areas were identified where improvements are either needed or should be considered to increase the safety of reactor operations

  13. IAEA programme on research reactor safety

    International Nuclear Information System (INIS)

    Alcala, F.; Di Meglio, A.F.

    1995-01-01

    This paper describes the IAEA programme on research reactor safety and includes the safety related areas of conversions to the use of low enriched uranium (LEU) fuel. The program is based on the IAEA statutory responsibilities as they apply to the requirements of over 320 research reactors operating around the world. The programme covers four major areas: (a) the development of safety documents; (b) safety missions to research reactor facilities; (c) support of research programmes on research reactor safety; (d) support of Technical Cooperation projects on research reactor safety issues. The demand for these activities by the IAEA member states has increased substantially in recent years especially in developing countries with increasing emphasis being placed on LEU conversion matters. In response to this demand, the IAEA has undertaken an extensive programme for each of the four areas above. (author)

  14. Fusion reactor safety

    International Nuclear Information System (INIS)

    1987-12-01

    Nuclear fusion could soon become a viable energy source. Work in plasma physics, fusion technology and fusion safety is progressing rapidly in a number of Member States and international collaboration continues on work aiming at the demonstration of fusion power generation. Safety of fusion reactors and technological and radiological aspects of waste management are important aspects in the development and design of fusion machines. In order to provide an international forum to review and discuss the status and the progress made since 1983 in programmes related to operational safety aspects of fusion reactors, their waste management and decommissioning concepts, the IAEA had organized the Technical Committee on ''Fusion Reactor Safety'' in Culham, 3-7 November 1986. All presentations of this meeting were divided into four sessions: 1. Statements on National-International Fusion Safety Programmes (5 papers); 2. Operation and System Safety (15 papers); 3. Waste Management and Decommissioning (5 papers); 4. Environmental Impacts (6 papers). A separate abstract was prepared for each of these 31 papers. Refs, figs, tabs

  15. Major Results from 1-Train Passive Safety System Tests for the SMART Design with the SMART-ITL Facility

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun-Sik; Bae, Hwang; Ryu, Sung-Uk; Jeon, Byong-Guk; Ruy, Hyobong; Kim, Woo-Shik; Byun, Sun-Joon; Shin, Yong-Cheol; Min, Kyoung-Ho; Yi, Sung-Jae [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    To satisfy the domestic and international needs for nuclear safety improvement after the Fukushima accident, an effort to improve its safety has been studied, and a Passive Safety System (PSS) for SMART has been designed. In addition, an Integral Test Loop for the SMART design (SMART-ITL, or FESTA) has been constructed and it finished its commissioning tests in 2012. Consequently, a set of Design Base Accident (DBA) scenarios have been simulated using SMARTITL. Recently, a test program to validate the performance of the SMART PSS was launched and its scaled-down test facility was additionally installed at the existing SMART-ITL facility. In this paper, the major results from the 1-train passive safety system validation tests with the SMARTITL facility will be summarized. The acquired data will be used to validate the safety analysis code and its related models, to evaluate the performance of SMART PSS, and to provide base data during the application phase of the SDA revision and construction licensing. In this paper, the major results from the validation tests of the SMART passive safety system using a 1-train test facility were summarized. They include a dozen of SMART PSS tests using 1-train SMART PSS tests. From the test results, it was estimated that the SMART PSS has sufficient cooling capability to deal with the SBLOCA scenario of SMART. During the SBLOCA scenario, in the CMT, the water layer inventory was well stratified thermally and the safety injection water was injected efficiently into the RPV from the initial period, and cools down the RCS properly.

  16. Evaluation of reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1960-04-15

    Although the operation of nuclear reactors has a remarkably good record of safety, the prevention of possible reactor accidents is one of the major factors that atomic planners have to contend with. At the same time, excessive caution may breed an attitude that hampers progress, either by resisting new development or by demanding unnecessarily elaborate and expensive precautions out of proportion to the actual hazards involved. The best course obviously is to determine the possible dangers and adopt adequate measures for their prevention, providing of course, for a reasonable margin of error in judging the hazards and the effectiveness of the measures. The greater the expert understanding and thoroughness with which this is done, the narrower need the margin be. This is the basic idea behind the evaluation of reactor safety

  17. INEL test reactor facility alarms: descriptions, technical specifications, and modification procedure

    International Nuclear Information System (INIS)

    Potash, L.M.; Boone, M.P.

    1980-04-01

    This report identifies standards, procedures, and practices which will affect any attempt to integrate or introduce human engineering principles into nuclear power plant alarm systems. Additional information concerning type of signal used, expected reaction, type of sensor, etc., is presented because of its relevance to future work on alarm system integration. The INEL test reactors were studied. Interviews were conducted with operators, designers, and management personnel. Additional information was obtained from available documentation. Only fire-alarm systems, and to a lesser extent, criticality alarms, have detailed industry-wide standards. One general standard has been written for control-room annunciators

  18. State of exposure control for workers engaging in radiation works and state of radioactive waste management in nuclear reactor facilities for test and research and nuclear reactor facilities at research and development stage, fiscal year 1995

    International Nuclear Information System (INIS)

    1996-01-01

    This is the summary of the reports submitted in fiscal year 1995 by the installers of the nuclear reactor facilities for test and research or at research and development stage, conforming to the related law. The individual dose equivalent of the workers engaging in radiation works in fiscal year 1995 was sufficiently lower than the prescribed limit in all reactor facilities. As for the released quantities of gaseous and liquid wastes, the radioactive substances in the air and water outside the monitor zones never exceeded the prescribed concentration limit in all reactor facilities. In the reactor facilities, for which the target values of release control have been determined, the values were less than the targets in all cases. The increase of stored radioactive solid waste decreased as the dismantling works of the reactor auxiliary system of the nuclear powered ship 'Mutsu' were finished in fiscal year 1994. As the amount of stored radioactive solid waste approaches the installed capacity, the preservation capacity of the existing waste preservation building was increased. (K.I.)

  19. IAEA safety standards for research reactors

    International Nuclear Information System (INIS)

    Abou Yehia, H.

    2007-01-01

    The general structure of the IAEA Safety Standards and the process for their development and revision are briefly presented and discussed together with the progress achieved in the development of Safety Standards for research reactor. These documents provide the safety requirements and the key technical recommendations to achieve enhanced safety. They are intended for use by all organizations involved in safety of research reactors and developed in a way that allows them to be incorporated into national laws and regulations. The author reviews the safety standards for research reactors and details their specificities. There are 4 published safety standards: 1) Safety assessment of research reactors and preparation of the safety analysis report (35-G1), 2) Safety in the utilization and modification of research reactors (35-G2), 3) Commissioning of research reactors (NS-G-4.1), and 4) Maintenance, periodic testing and inspection of research reactors (NS-G-4.2). There 5 draft safety standards: 1) Operational limits and conditions and operating procedures for research reactors (DS261), 2) The operating organization and the recruitment, training and qualification of personnel for research reactors (DS325), 3) Radiation protection and radioactive waste management in the design and operation of research reactors (DS340), 4) Core management and fuel handling at research reactors (DS350), and 5) Grading the application of safety requirements for research reactors (DS351). There are 2 planned safety standards, one concerning the ageing management for research reactor and the second deals with the control and instrumentation of research reactors

  20. Safety of research reactors (Design and Operation)

    International Nuclear Information System (INIS)

    Dirar, H. M.

    2012-06-01

    The primary objective of this thesis is to conduct a comprehensive up-to-date literature review on the current status of safety of research reactor both in design and operation providing the future trends in safety of research reactors. Data and technical information of variety selected historical research reactors were thoroughly reviewed and evaluated, furthermore illustrations of the material of fuel, control rods, shielding, moderators and coolants used were discussed. Insight study of some historical research reactors was carried with considering sample cases such as Chicago Pile-1, F-1 reactor, Chalk River Laboratories,. The National Research Experimental Reactor and others. The current status of research reactors and their geographical distribution, reactor category and utilization is also covered. Examples of some recent advanced reactors were studied like safety barriers of HANARO of Korea including safety doors of the hall and building entrance and finger print identification which prevent the reactor from sabotage. On the basis of the results of this research, it is apparent that a high quality of safety of nuclear reactors can be attained by achieving enough robust construction, designing components of high levels of efficiency, replacing the compounds of the reactor in order to avoid corrosion and degradation with age, coupled with experienced scientists and technical staffs to operate nuclear research facilities.(Author)

  1. Safety of research reactors - A regulator's perspective

    International Nuclear Information System (INIS)

    Rahman, M.S.

    2001-01-01

    Due to historical reasons research reactors have received less regulatory attention in the world than nuclear power plants. This has given rise to several safety issues which, if not addressed immediately, may result in an undesirable situation. However, in Pakistan, research reactors and power reactors have received due attention from the regulatory authority. The Pakistan Research Reactor-1 has been under regulatory surveillance since 1965, the year of its commissioning. The second reactor has also undergone all the safety reviews and checks mandated by the licensing procedures. A brief description of the regulatory framework, the several safety reviews carried out have been briefly described in this paper. Significant activities of the regulatory authority have also been described in verifying the safety of research reactors in Pakistan along with the future activities. The views of the Pakistani regulatory authority on the specific issues identified by the IAEA have been presented along with specific recommendations to the IAEA. We are of the opinion that there are more Member States operating nuclear research reactors than nuclear power plants. Therefore, there should be more emphasis on the research reactor safety, which somehow has not been the case. In several recommendations made to the IAEA on the specific safety issues the emphasis has been, in general, to have a similar documentation and approach for maintaining and verifying operational safety at research reactors as is currently available for nuclear power reactors and may be planned for nuclear fuel cycle facilities. (author)

  2. Exercising Synergy of Safeguards Safety and Security at Facility Level of the GA Siwabessy Multi-Purpose Reactor, Indonesia

    International Nuclear Information System (INIS)

    Susilowati, E.

    2015-01-01

    Safeguards, safety and security (3Ss) constitute as essential elements for successful development of nuclear technology in the life time of nuclear installation. All 3Ss need to be coordinated due workers, the public and the environment require protection from plant malfunction, human error, malicious acts and proliferation of nuclear materials and technologies. Then the importance of the 3Ss was deemed valuable, particularly to a country having willingness to expand to nuclear power reactor such as Indonesia that in the near future plans to build small experimental power reactor. This paper is aimed to discuss synergy among safeguards, safety and security which will have opportunity been exercising at the GA Siwabessy Reactor (RSG-GAS), Indonesia. Synergy among safeguards, safety and security offers much opportunity for cost savings and enhance efficiency. Discussion is carried out by first investigating common values and conflicts exist among 3S. Up to now each of them was accomplished separately by different division and using different equipment due lack of coordination among them. The objective of this exercise is to develop more efficient and effective 3Ss infrastructures and also to support skill and knowledge of human resources. Benefitting from synergy between safeguards and security such as management of nuclear material and non proliferation; safeguards and safety such as management of nuclear material and waste management; safety and security such as prevent radiological release and also tension among them if any are discussed. It is expected that outcome of this exercise will able to develop a role model of infrastructures to the up-coming small experimental power reactor in Indonesia. (author)

  3. Relationship between general safety requirements and safety culture in the improvement of safe operation of I.N.R. TRIGA reactor facilities

    International Nuclear Information System (INIS)

    Ciocanescu, M.; Preda, M.; Chiritescu, M.; Dumitru, M.

    1996-01-01

    Acquiring of the basic principles of ''safety culture'' by a large number of profesionals in the nuclear field drew the attention of the decision factors in the INR managerial structure, who decided to promote certain practical actions at each level in order to improve nuclear safety. Starting from the ''Republican Standards for Nuclear Safety'' issued by CSEN in 1975, where general safety criteria are defined for nuclear reactors and NPPs, the specialists at the TRIGA reactor originated and implemented a coherent and secure system to ensure nuclear safety over all steps of nuclear activities: research, conception, execution, commissioning and operation. This system has been continuosly corrected so that now it is completely integrated in a modern safety system. The paper presents the way in which a modern system for nuclear safety at the TRIGA reactor has been implemented and developed, in accordance to specific criteria and requirements imposed by related National Regulations and with the principles of safety culture. Starting from the definition of specific responsabilities, there are presented the internal stipulations and practical actions at all levels in order to enhance nuclear safety. (orig.)

  4. Reactor safety facility

    International Nuclear Information System (INIS)

    Takii, Taichi; Suzuki, Hiroaki; Murase, Michio.

    1995-01-01

    An opening of a bent tube in communication with an upper dry well and a pressure suppression chamber is disposed higher by not less than 2m than the lower end of the pressure vessel, and an opening of an upper and lower dry well communication pipe is disposed lower than the opening of the bent tube, further, a flow channel in communication with the upper and lower portions of a support of the pressure vessel is disposed. A flow channel for cooling water or steams is formed between the outer wall of the pressure vessel and a heat insulation wall, and an opening/closing mechanism which is closed during normal operation and opened upon occurrence of accident is disposed at least to one place of the channel. Cooling water supplied to the upper dry well flows to the lower dry well from the opening of the upper and lower dry well communication pipe, accumulates in the upper dry well and reaches at a height of the opening of the bent tube and flows into the pressure suppression chamber only after submerging the pressure vessel. Thus, the water level of the pressure suppression chamber is not elevated for cooling the pressure vessel, and an external water source is not necessary. (N.H.)

  5. Irradiation facilitates at the advanced test reactor

    International Nuclear Information System (INIS)

    Grover, Blaine S.

    2006-01-01

    The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC - formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950's with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world's data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens. The paper has the following contents: ATR description and capabilities; ATR operations, quality and safety requirements; Static capsule experiments; Lead experiments; Irradiation test vehicle; In-pile loop experiments; Gas test loop; Future testing; Support facilities at RTC; Conclusions. To summarize, the ATR has a long history in fuel and material irradiations, and will be fulfilling a critical role in the future fuel and material testing necessary to develop the next generation reactor systems and advanced fuel cycles. The

  6. Perspectives on reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)

    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.

  7. Perspectives on reactor safety

    International Nuclear Information System (INIS)

    Haskin, F.E.

    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

  8. RB research reactor Safety Report

    International Nuclear Information System (INIS)

    Sotic, O.; Pesic, M.; Vranic, S.

    1979-04-01

    This RB reactor safety report is a revised and improved version of the Safety report written in 1962. It contains descriptions of: reactor building, reactor hall, control room, laboratories, reactor components, reactor control system, heavy water loop, neutron source, safety system, dosimetry system, alarm system, neutron converter, experimental channels. Safety aspects of the reactor operation include analyses of accident causes, errors during operation, measures for preventing uncontrolled activity changes, analysis of the maximum possible accident in case of different core configurations with natural uranium, slightly and highly enriched fuel; influence of possible seismic events

  9. Preliminary Feasibility, Design, and Hazard Analysis of a Boiling Water Test Loop Within the Idaho National Laboratory Advanced Test Reactor National Scientific User Facility

    International Nuclear Information System (INIS)

    Gerstner, Douglas M.

    2009-01-01

    The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The ATR and its support facilities are located at the Idaho National Laboratory (INL). A Boiling Water Test Loop (BWTL) is being designed for one of the irradiation test positions within the. The objective of the new loop will be to simulate boiling water reactor (BWR) conditions to support clad corrosion and related reactor material testing. Further it will accommodate power ramping tests of candidate high burn-up fuels and fuel pins/rods for the commercial BWR utilities. The BWTL will be much like the pressurized water loops already in service in 5 of the 9 'flux traps' (region of enhanced neutron flux) in the ATR. The loop coolant will be isolated from the primary coolant system so that the loop's temperature, pressure, flow rate, and water chemistry can be independently controlled. This paper presents the proposed general design of the in-core and auxiliary BWTL systems; the preliminary results of the neutronics and thermal hydraulics analyses; and the preliminary hazard analysis for safe normal and transient BWTL and ATR operation

  10. Meltdown reactor core cooling facility

    International Nuclear Information System (INIS)

    Matsuoka, Tsuyoshi.

    1992-01-01

    The meltdown reactor core cooling facility comprises a meltdown reactor core cooling tank, a cooling water storage tank situates at a position higher than the meltdown reactor core cooling tank, an upper pipeline connecting the upper portions of the both of the tanks and a lower pipeline connecting the lower portions of them. Upon occurrence of reactor core meltdown, a high temperature meltdown reactor core is dropped on the cooling tank to partially melt the tank and form a hole, from which cooling water is flown out. Since the water source of the cooling water is the cooling water storage tank, a great amount of cooling water is further dropped and supplied and the reactor core is submerged and cooled by natural convection for a long period of time. Further, when the lump of the meltdown reactor core is small and the perforated hole of the meltdown reactor cooling tank is small, cooling water is boiled by the high temperature lump intruding into the meltdown reactor core cooling tank and blown out from the upper pipeline to the cooling water storage tank to supply cooling water from the lower pipeline to the meltdown reactor core cooling tank. Since it is constituted only with simple static facilities, the facility can be simplified to attain improvement of reliability. (N.H.)

  11. Power reactor core safety research

    International Nuclear Information System (INIS)

    Rim, C.S.; Kim, W.C.; Shon, D.S.; Kim, J.

    1981-01-01

    As a part of nuclear safety research program, a project was launched to develop a model to predict fuel failure, to produce the data required for the localizaton of fuel design and fabrication technology, to establish safety limits for regulation of nuclear power plants and to develop reactor operation method to minimize fuel failure through the study of fuel failure mechanisms. During 1980, the first year of this project, various fuel failure mechanisms were analyzed, an experimental method for out-of-pile tests to study the stress corrosion cracking (SCC) behaviour of Zircaloy cladding underiodine environment was established, and characteristics of PWR and CANDU Zircaloy specimens were examined. Also developed during 1980 were the methods and correlations to evaluate fuel failures in the reactor core based on operating data from power reactors

  12. Reactor safety systems

    International Nuclear Information System (INIS)

    Kafka, P.

    1975-01-01

    The spectrum of possible accidents may become characterized by the 'maximum credible accident', which will/will not happen. Similary, the performance of safety systems in a multitude of situations is sometimes simplified to 'the emergency system will/will not work' or even 'reactors are/ are not safe'. In assessing safety, one must avoid this fallacy of reducing a complicated situation to the simple black-and-white picture of yes/no. Similarly, there is a natural tendency continually to improve the safety of a system to assure that it is 'safe enough'. Any system can be made safer and there is usually some additional cost. It is important to balance the increased safety against the increased costs. (orig.) [de

  13. New Sensors for In-Pile Temperature Detection at the Advanced Test Reactor National Scientific User Facility

    International Nuclear Information System (INIS)

    Rempe, J.L.; Knudson, D.L.; Daw, J.E.; Condie, K.G.; Wilkins, S. Curtis

    2009-01-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. As a user facility, the ATR is supporting new users from universities, laboratories, and industry, as they conduct basic and applied nuclear research and development to advance the nation's energy security needs. A key component of the ATR NSUF effort is to develop and evaluate new in-pile instrumentation techniques that are capable of providing measurements of key parameters during irradiation. This paper describes the strategy for determining what instrumentation is needed and the program for developing new or enhanced sensors that can address these needs. Accomplishments from this program are illustrated by describing new sensors now available and under development for in-pile detection of temperature at various irradiation locations in the ATR.

  14. THAI test facility for experimental research on hydrogen and fission product behaviour in light water reactor containments

    Energy Technology Data Exchange (ETDEWEB)

    Gupta, S., E-mail: gupta@becker-technologies.com [Becker Technologies GmbH, Koelner Strasse 6, 65760 Eschborn (Germany); Schmidt, E.; Laufenberg, B. von; Freitag, M.; Poss, G. [Becker Technologies GmbH, Koelner Strasse 6, 65760 Eschborn (Germany); Funke, F. [AREVA GmbH, P.O. Box 1109, 91001 Erlangen (Germany); Weber, G. [Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH, Forschungszentrum, Boltzmannstraße 14, 85748 Garching (Germany)

    2015-12-01

    Highlights: • Large scale facility for investigating representative LWR severe accident scenarios. • Coupled effect tests in the field of thermal-hydraulics, hydrogen, aerosol and iodine. • Measurement techniques improved and adapted for severe accident conditions. • Testing of passive mitigation systems (e.g. PAR) under accident conditions. • THAI data application for validation and development of CFD and LP codes. - Abstract: The test facility THAI (thermal-hydraulics, hydrogen, aerosol, and iodine) aims at addressing open questions concerning gas distribution, behaviour of hydrogen, iodine and aerosols in the containment of light water reactors during severe accidents. Main component of the facility is a 60 m{sup 3} stainless steel vessel, 9.2 m high and 3.2 m in diameter, with exchangeable internals for multi-compartment investigations. The maximal design pressure of the vessel is 14 bar which allows H{sub 2} combustion experiments at a severe accident relevant H{sub 2} concentration level. The facility is approved for the use of low-level radiotracer I-123 which enables the measurement of time resolved iodine behaviour. The THAI test facility allows investigating various accident scenarios, ranging from turbulent free convection to stagnant stratified containment atmospheres and can be combined with simultaneous use of hydrogen, iodine and aerosol issues. THAI experimental research also covers investigations related to mitigation systems employed in light water reactor containments by performing experiments on, e.g. pressure suppression pool hydrodynamics, performance behaviour of passive autocatalytic recombiners, and spray interaction with hydrogen–steam–air flames in phenomenon orientated and coupled-effects experiments. The THAI experimental data have been widely used for the validation and further development of Lumped Parameter and Computational Fluid Dynamics codes with 3D capabilities, e.g. International Standard Problems ISP-47 (thermal

  15. Safety of Research Reactors. Safety Requirements

    International Nuclear Information System (INIS)

    2010-01-01

    The main objective of this Safety Requirements publication is to provide a basis for safety and a basis for safety assessment for all stages in the lifetime of a research reactor. Another objective is to establish requirements on aspects relating to regulatory control, the management of safety, site evaluation, design, operation and decommissioning. Technical and administrative requirements for the safety of research reactors are established in accordance with these objectives. This Safety Requirements publication is intended for use by organizations engaged in the site evaluation, design, manufacturing, construction, operation and decommissioning of research reactors as well as by regulatory bodies

  16. Impurity Control Test Facility (ICTF) for the study of fusion reactor plasma/edge materials interactions

    International Nuclear Information System (INIS)

    Brooks, J.N.; Mattas, R.F.; Ehst, D.A.; Boley, C.D.; Hershkowitz, N.

    1984-05-01

    A test facility for investigating many of the impurity control issues associated with the interactions of materials with the plasma edge is outlined. Analysis indicates that the plasma edge conditions expected in TFCX, INTOR, etc. can be readily produced at the end cells of an rf stabilized mirror, similar in some respects to the Phaedrus device at the University of Wisconsin. A steady-state, Impurity Control Test Facility (ICTF) based on such a mirror device is expected to produce a plasma with typical parameters of n/sub e/ approx. 3 x 10 18 m -3 , T/sub e/ = 50 eV, and T/sub i/ = 100 eV at each end cell. A heat load of approx. 2 MW/m 2 over areas of approx. 1600 cm 2 could be produced at each end with 800 kW of ICRH power. These conditions would provide a unique capability for examining issues such as erosion/redeposition behavior, properties of redeposited materials, high recycling regimes, plasma edge operating limits for high-Z materials, and particle pumping efficiencies for limiter and divertor designs

  17. Potential of high speed x-ray cinematography as a fuel motion diagnostic for safety test facilities

    International Nuclear Information System (INIS)

    Stalker, K.T.; Choate, L.M.; Posey, L.D.

    1979-01-01

    Experiments have been performed which indicate the feasibility of using X-ray cinematography as a diagnostic tool for monitoring fuel motion in large pin bundle advanced reactor safety tests. This capability was demonstrated by imaging motion in a 37-pin bundle of simulated fuel elements at a data rate of 400 pictures per second using an active detector system coupled with a 10 MeV accelerator. 5 refs

  18. PWR-related integral safety experiments in the PKL 111 test facility SBLOCA under beyond-design-basis accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Weber, P.; Umminger, K.J.; Schoen, B. [Siemens AG Power Generation Group (KWU), Erlangen (France)

    1995-09-01

    The thermal hydraulic behavior of a PWR during beyond-design-basis accident scenarios is of vital interest for the verification and optimization of accident management procedures. Within the scope of the German reactor safety research program experiments were performed in the volumetrically scaled PKL 111 test facility by Siemens/KWU. This highly instrumented test rig simulates a KWU-design PWR (1300 MWe). In particular, the latest tests performed related to a SBLOCA with additional system failures, e.g. nitrogen entering the primary system. In the case of a SBLOCA, it is the goal of the operator to put the plant in a condition where the decay heat can be removed first using the low pressure emergency core cooling system and then the residual heat removal system. The experimental investigation presented assumed the following beyond-design-basis accident conditions: 0.5% break in a cold leg, 2 of 4 steam generators (SGs) isolated on the secondary side (feedwater- and steam line-valves closed), filled with steam on the primary side, cooldown of the primary system using the remaining two steam generators, high pressure injection system only in the two loops with intact steam generators, if possible no operator actions to reach the conditions for residual heat removal system activation. Furthermore, it was postulated that 2 of the 4 hot leg accumulators had a reduced initial water inventory (increased nitrogen inventory), allowing nitrogen to enter the primary system at a pressure of 15 bar and nearly preventing the heat transfer in the SGs ({open_quotes}passivating{close_quotes} U-tubes). Due to this the heat transfer regime in the intact steam generators changed remarkably. The primary system showed self-regulating system effects and heat transfer improved again (reflux-condenser mode in the U-tube inlet region).

  19. Safety assesment necessary in selecting the technologies for partial decommissioning of nuclear facilities. Application to research reactors

    International Nuclear Information System (INIS)

    Niculae, O.; Stan, C.; Vladescu, G.

    2005-01-01

    The main goal of this work is identification and evaluation of safety indicators - quantities used in monitoring the safety assurance during decommissioning processes in nuclear facilities identification of safety indicators is made on basis of qualitative and quantitative analysis, effected for both normal decommissioning, as well as in case of foreseen event occurrence. The safety indicators form an integrated system which can be represented by a pyramidal structural with the following levels (in increasing complexity order): specific indicators, strategic indicators, overall indicators, safety closure. This work suggests that evaluation of safety assurance level during the conduct of a decommissioning process to be based on the overall analysis of the set of indicators emphasizing not only the evaluation of individual safety indicators but also the interdependencies between them. The evaluation method is based on the 'step-by-step' principle. The evaluation was carried-out either directly or by means of dedicated evaluation forms which cover both quantitative and qualitative aspects of the analysis. At the some time identified are the adequate protection measures for the personnel implied in decommissioning, as well as for population and environment. The paper present also technologies adequate in the decommissioning. (authors)

  20. Safety review and stress test of the uranium enrichment facility Gronau; Compact. Sicherheitsueberpruefung und Stresstest der Urananreicherungsanlage Gronau

    Energy Technology Data Exchange (ETDEWEB)

    Kleiboemer, Burkhard [URENCO Deutschland GmbH, Gronau (Germany). Ueberwachung

    2013-07-01

    As a consequence of the earthquake and tsunami in Japan in 2011 and the subsequent catastrophic destruction of the nuclear power plant in Fukushima URENCO has agreed to perform the safety check of the uranium enrichment facility Gronau (UAG) two years earlier than planned. The review was started in May 2011. The contribution deals with the requirements for the safety analysis and the questionnaire for the stress test, the methodology for realization of the safety check and the results of the investigations performed by URENCO and experts of ESK.

  1. Probabilistic methods of optimization of scheduled tests for heat equipment of safety systems of reactor at full power

    International Nuclear Information System (INIS)

    Bilej, D.V.; Fridman, N.A.; Kolykhanov, V.N.; Skalozubov, V.I.

    2004-01-01

    This article generalises the basic results of a long-term teamwork with respect to a scientific and technical substantiation of perfection of the regulations of safe operation power units with VVER. This perfection is concerning a periodicity and volumes of tests of safety systems when a reactor works at full power. The article shows that the application of the probabilistic approaches connected to minimisation of a risk criterion function is an effective methodical base for the optimisation. For certain safety systems of serial power units with VVER 1000 the results of calculated substantiations are presented

  2. Nuclear fuel cycle facilities, laboratories, irradiators, particle accelerators, under-decommissioning reactors and radioactive waste management facilities safety. Lessons learned from events notified between 2005 and 2008

    International Nuclear Information System (INIS)

    2001-01-01

    Maintaining high levels of safety in nuclear facilities requires constant vigilance by everyone involved, especially by plant operators who are first and foremost responsible for safety in their facilities. Safety can never be taken for granted; constant efforts must be made to improve it, by taking new knowledge and available operating feedback into account. In this respect, a substantial part of operating feedback is made up of lessons learned from analysing events, incidents or accidents occurring in France or in similar facilities abroad. To encourage the diffusion of operating feedback, IRSN has produced a report concerning events notified to the Nuclear Safety Authority (ASN) by operators of LUDD facilities between 2005 and 2008. The main objective is to make general lessons for safety in this type of facility available based on a cross-disciplinary analysis of notified events and noted evolution trends. IRSN has had tools for managing information concerning events occurring in France and abroad for many years. These tools are used to analyse the events in order to take into account the relevant lessons learned in the safety assessments performed on behalf of ASN and also to define study and research programmes to maintain its expertise and expand its knowledge. The report has 4 sections: - the first section (chapters 2 to 4) presents the LUDD facilities so that the facilities themselves, their diversity and the main associated risks can be better understood. It also includes a brief reminder of plant operator obligations in notifying events and describes the database used by the Institute to manage the data relating to the notified events; - the second section (chapter 5) summarises the main changes noted in the events notified to ASN during 2005 to 2008 and provides an overall assessment of the consequences of these events for the environment, the population and the workers; - the third section (chapter 6) describes significant events occurring in France

  3. The SPHINX reactor for engineering tests

    International Nuclear Information System (INIS)

    Adamov, E.O.; Artamkin, K.N.; Bovin, A.P.; Bulkin, Y.M.; Kartashev, E.F.; Korneev, A.A.; Stenbok, I.A.; Terekhov, A.S.; Khmel'Shehikov, V.V.; Cherkashov, Y.M.

    1990-01-01

    A research reactor known as SPHINX is under development in the USSR. The reactor will be used mainly to carry out tests on mock-up power reactor fuel assemblies under close-to-normal parameters in experimental loop channels installed in the core and reflector of the reactor, as well as to test samples of structural materials in ampoule and loop channels. The SPHINX reactor is a channel-type reactor with light-water coolant and moderator. Maximum achievable neutron flux density in the experimental channels (cell composition 50% Fe, 50% H 2 O) is 1.1 X 10 15 neutrons/cm 2 · s for fast neutrons (E > 0.1 MeV) and 1.7 X 10 15 for thermal neutrons at a reactor power of 200 MW. The design concepts used represent a further development of the technical features which have met with approval in the MR and MIR channel-type engineering test reactors currently in use in the USSR. The 'in-pond channel' construction makes the facility flexible and eases the carrying out of experimental work while keeping discharges of radioactivity into the environment to a low level. The reactor and all associated buildings and constructions conform to modern radiation safety and environmental protection requirements

  4. Aspects of nuclear reactor safety

    International Nuclear Information System (INIS)

    Hardt, P. von der; Rottger, H.

    1980-01-01

    The Colloquium on 'Irradiation Tests for Reactor Safety Programmes' has been organised by JRC Petten in order to determine the present state of technology in the field. The role of research and test reactors for studies of structural material and fuel elements under transient and off-normal conditions was to be explained. The Colloquium has been attended by 110 participants from outside and inside Europe. 27 papers were presented covering the major ongoing projects in Japan, the United States, and in Europe, and elaborating in particular: - design rationale and layout of safety irradiation experiments; - design, manufacture, and performance of irradiation equipment with particular attention to generation and control of transient conditions, fast response in-pile instrumentation and its out-of-pile data retrieval; - post-irradiation evaluation; - results and analytical support

  5. Refurbishment and safety upgradation of research reactor Cirus

    International Nuclear Information System (INIS)

    Marik, S.K.; Rao, D.V.H.; Bhatnagar, A.; Pant, R.C.; Tikku, A.C.; Sankar, S.

    2006-01-01

    Cirus, a 40 MW t, vertical tank type research reactor, having wide range of research facilities, was commissioned in the year 1960. This research reactor, situated at Mumbai, India has been operated and utilized extensively for isotope production, material testing and neutron beam research for nearly four decades. With a view to assess the residual life of the reactor, detailed ageing studies were carried out during the early 1990s. Based on these studies, refurbishment of Cirus for its life extension was taken up. During refurbishment, additional safety features were incorporated in various systems to qualify them for the current safety standards. This paper gives the details of the operating experiences, utilization of the reactor along with methodologies followed for carrying out detailed ageing studies, refurbishment and safety upgradation for its life extension

  6. Nuclear reactors safety issues

    International Nuclear Information System (INIS)

    Barre, Francois; Seiler, Nathalie

    2008-01-01

    Full text of publication follows: Since the seventies, economic incentives have led the utilities to drive a permanent evolution of the light water reactor (LWR). The evolution deals with the reactor designs as well as the way to operate them in a more flexible manner. It is for instance related to the fuel technologies and management. On the one hand, the technologies are in continuous evolution, such as the fuel pellets (MOX, Gd fuel, or Cr doped fuels..) as well as advanced cladding materials (M5 TM , MDA or ZIRLO). On the other hand, the fuel management is also subject to continuous evolution in particular in terms of increasing the level of burn-up, the reactor (core) power, the enrichment, as well as the duration of reactor cycles. For instance, in a few years in France, the burn-up has raised beyond the value of 39 GWj/t, initially authorized up to 52 GWj/t for the UO 2 fuel. In the near future, utilities foreseen to reach fuel burn-up of 60 GWj/t for MOX fuel and 70 GWj/t for UO 2 fuel. Furthermore, the future reactor of fourth generation will use new fuels of advanced conception. Furthermore with the objective of improving the safety margins, methods and calculation tools used by the utilities in the elaboration of their safety demonstrations submitted to the Safety Authority, are in movement. The margin evaluation methodologies often consist of a calculation chain of best-estimate multi-field simulations (e.g. various codes being coupled to simulate in a realistic way the evolution of the thermohydraulic, neutronic and mechanic state of the reactor). The statistical methods are more and more sophisticated and the computer codes are integrating ever-complex physical models (e.g. three-dimensional at fine scale). Following this evolution, the Institute of Radioprotection and Nuclear Safety (IRSN), whose one of the roles is to examine the safety records and to rend a technical expertise, considers the necessity of reevaluating the safety issues for advanced

  7. Heat removal tests for pressurized water reactor containment spray by largescale facility

    International Nuclear Information System (INIS)

    Motoki, Y.; Hashimoto, K.; Kitani, S.; Naritomi, M.; Nishio, G.; Tanaka, M.

    1983-01-01

    Heat removal tests for pressurized water reactor (PWR) containment spray were carried out to investigate effectiveness of the depressurization by Japan Atomic Energy Research Institute model containment (7-m diameter, 20 m high, and 708-m 3 volume) with PWR spray nozzles. The depressurization rate is influenced by the spray heat transfer efficiency and the containment wall surface heat transfer coefficient. The overall spray heat transfer efficiency was investigated with respect to spray flow rate, weight ratio of steam/air, and spray height. The spray droplet heat transfer efficiency was investigated whether the overlapping of spray patterns gives effect or not. The effect was not detectable in the range of large value of steam/air, however, it was better in the range of small value of it. The experimental results were compared with the calculated results by computer code CONTEMPT-LT/022. The overall spray heat transfer efficiency was almost 100% in the containment pressure, ranging from 2.5 to 0.9 kg/cm 2 X G, so that the code was useful on the prediction of the thermal hydraulic behavior of containment atmosphere in a PWR accident condition

  8. The second eddy current testing of zircaloy tube samples from the OECD Halden reactor project at Reactor Fuel Examination Facility, Tokai, JAERI

    International Nuclear Information System (INIS)

    Ohwada, Isao; Nishino, Yasuharu

    1986-07-01

    The Reactor Fuel Examination Facility in Tokai/JAERI (Japan Atomic Energy Research Institute) joined to the second round robin programme on eddy current test of the Halden/IFE. In the programme, two zircaloy tube samples with some artificial defects were provided for measurements. To clarify the locations in axial and azimuthal directions, types and dimensions of the provided artificial defects, measured signals from eddy current test were analysed in comparison with the known defects on the calibration tube. As a result, fourteen defects were determined from the measurements. Then, the location, the type and the relative dimension of them were also revealed. The results of those eddy current test are described in this paper. (author)

  9. Risk management activities at the DOE Class A reactor facilities

    International Nuclear Information System (INIS)

    Sharp, D.A.; Hill, D.J.; Linn, M.A.; Atkinson, S.A.; Hu, J.P.

    1993-01-01

    The probabilistic risk assessment (PRA) and risk management group of the Association for Excellence in Reactor Operation (AERO) develops risk management initiatives and standards to improve operation and increase safety of the DOE Class A reactor facilities. Principal risk management applications that have been implemented at each facility are reviewed. The status of a program to develop guidelines for risk management programs at reactor facilities is presented

  10. Light water reactor safety research project

    International Nuclear Information System (INIS)

    Markoczy, G.; Aksan, S.N.; Behringer, K.; Prodan, M.; Stierli, F.; Ullrich, G.

    1980-07-01

    The research and development activities for the safety of Light Water Power Reactors carried out 1979 at the Swiss Federal Institute for Reactor Research are described. Considerations concerning the necessity, objectives and size of the Safety Research Project are presented, followed by a detailed discussion of the activities in the five tasks of the program, covering fracture mechanics and nondestructive testing, thermal-hydraulics, reactor noise analysis and pressure vessel steel surveillance. (Auth.)

  11. Research on the reactor physics and reactor safety of VVER reactors. AER Symposium 2016

    Energy Technology Data Exchange (ETDEWEB)

    Kliem, S.

    2017-09-15

    The selected paperscan be attributed to the following main subjects: Reactor start-up tests and use of corresponding data for code validation, code development and application, approaches for safety analyses, closure of nuclear fuel cycle, prospective reactor concepts.

  12. THORS: a high-temperature sodium test facility rated at 2.0 MW

    International Nuclear Information System (INIS)

    Gnadt, P.A.; Anderson, A.H.; Clapp, N.E.; Montgomery, B.H.; Collins, C.W.; Stulting, R.D.

    1979-01-01

    The Thermal--Hydraulic Out-of-Reactor Safety (THORS) facility at Oak Ridge Naitonal Laboratory (ORNL) is a high-temperature sodium test facility operated for the United States Breeder Reactor Safety Program. The facility is primarily used for testing large simulated Liquid-Metal Fast Breeder Reactor (LMFBR) fuel subassemblies. The facility has recently been upgraded to provide a 2.0-MW test bundle power input and heat removal capability. A new test section, which will be capable of operating at 980 0 C and which will accommodate a 217-pin bundle, has also been added. A 61-pin bundle is currently under test in the facility. A description of the test facility is presented, along with a brief summary of the 8-year operating history of this safety-related test facility

  13. Reactor containing facility

    International Nuclear Information System (INIS)

    Akagawa, Katsuhiko.

    1992-01-01

    A cooling space having a predetermined capacity is formed between a reactor container and concrete walls. A circulation loop disposed to the outside of the concrete walls is connected to the top and the bottom of the cooling space. The circulation loop has a circulation pump and a heat exchanger, and a cooling water supply pipe is connected to the upstream of the circulation pump for introducing cooling water from the outside. Upon occurrence of loss of coolant accident, cooling water is introduced from the cooling water supply pipe to the cooling space between the reactor container and the concrete walls after shut-down of the reactor operation. Then, cooling water is circulated while being cooled by the heat exchanger, to cool the reactor container by cooling water flown in the cooling space. This can cool the reactor container in a short period of time upon occurrence of the loss of coolant accident. Accordingly, a repairing operation for a ruptured portion can be conducted rapidly. (I.N.)

  14. Safety of nuclear power reactors

    International Nuclear Information System (INIS)

    MacPherson, H.G.

    1982-01-01

    Safety is the major public issue to be resolved or accommodated if nuclear power is to have a future. Probabilistic Risk Analysis (PRA) of accidental releases of low-level radiation, the spread and activity of radiation in populated areas, and the impacts on public health from exposure evolved from the earlier Rasmussen Reactor Safety Study. Applications of the PRA technique have identified design peculiarities in specific reactors, thus increasing reactor safety and establishing a quide for evaluating reactor regulations. The Nuclear Regulatory Commission and reactor vendors must share with utilities the responsibility for reactor safety in the US and for providing reasonable assurance to the public. This entails persuasive public education and information that with safety a top priority, changes now being made in light water reactor hardware and operations will be adequate. 17 references, 2 figures, 2 tables

  15. An experimental study on the thermal-hydraulic phenomena in the Hybrid Safety Injection Tank using a separate effect test facility

    International Nuclear Information System (INIS)

    Ryu, Sung Uk; Ryu, Hyobong; Park, Hyun-Sik; Yi, Sung-Jae

    2016-01-01

    Highlights: • The experimental study on the pressure balancing between the Hybrid SIT and PZR. • The effects of different variables affecting the pressure balancing are investigated. • A sensitivity analysis on the pressure variations of the Hybrid SIT. - Abstract: This paper reports an experimental research for investigating thermal hydraulic phenomena of Hybrid Safety Injection Tank (Hybrid SIT) using a separate effect test facility in Korea Atomic Energy Research Institute (KAERI). The Hybrid SIT is a passive safety injection system that enables the safety injection water to be injected into the reactor pressure vessel throughout all operating pressures by connecting the top of the SIT and the pressurizer (PZR). The separate effect test (SET) facility of Hybrid SIT, which is designed based on the APR+ power plant, comprises a PZR, Hybrid SIT, pressure balancing line (PBL), injection line (IL), nitrogen gas line, and refueling water tank (RWT). Furthermore, the pressure loss range of the SET facility was analyzed and compared with that of the reference nuclear power plant. In this research, a condition for balancing the pressure between the Hybrid SIT and PZR is examined and the effects of different variables affecting the pressure balancing, which are flow rate, injection velocity of steam and initial water level, are also investigated. The condition for balancing the pressure between the Hybrid SIT and PZR was derived theoretically from a pressure network for the Hybrid SIT, pressurizer, and reactor pressure vessel. Additionally, a sensitivity analysis as a theoretical approach was conducted on the pressure variations in relation to the rate of steam condensation inside the Hybrid SIT. The results showed that pressure of the Hybrid SIT was predominantly determined by the rate of steam condensation. The results showed that if the rate of condensation increased or decreased by 10%, the Hybrid SIT pressure at the pressure balancing point decreased or

  16. Reactor container facility

    International Nuclear Information System (INIS)

    Saito, Takashi; Nagasaka, Hideo.

    1990-01-01

    A dry-well pool for spontaneously circulating stored pool water and a suppression pool for flooding a pressure vessel by feeding water, when required, to a flooding gap by means of spontaneous falling upto the flooding position, thereby flooding the pressure vessel are contained at the inside of a reactor container. Thus, when loss of coolant accidents such as caused by main pipe rupture accidents should happen, pool water in the suppression pool is supplied to the flooding gap by spontaneously falling. Further, if the flooding water uprises exceeding a predetermined level, the flooding gap is in communication with the dry-well pool at the upper and the lower portions respectively. Accordingly, flooding water at high temperature heated by the after-heat of the reactor core is returned again into the flooding gap to cool the reactor core repeatedly. (T.M.)

  17. Simulator for materials testing reactors

    International Nuclear Information System (INIS)

    Takemoto, Noriyuki; Sugaya, Naoto; Ohtsuka, Kaoru; Hanakawa, Hiroki; Onuma, Yuichi; Hosokawa, Jinsaku; Hori, Naohiko; Kaminaga, Masanori; Tamura, Kazuo; Hotta, Kohji; Ishitsuka, Tatsuo

    2013-06-01

    A real-time simulator for both reactor and irradiation facilities of a materials testing reactor, “Simulator of Materials Testing Reactors”, was developed for understanding reactor behavior and operational training in order to utilize it for nuclear human resource development and to promote partnership with developing countries which have a plan to introduce nuclear power plant. The simulator is designed based on the JMTR (Japan Materials Testing Reactor), and it simulates operation, irradiation tests and various kinds of anticipated operational transients and accident conditions caused by the reactor and irradiation facilities. The development of the simulator was sponsored by the Japanese government as one of the specialized projects of advanced research infrastructure in order to promote basic as well as applied researches. This report summarizes the simulation components, hardware specification and operation procedure of the simulator. (author)

  18. Overview of fusion reactor safety

    International Nuclear Information System (INIS)

    Cohen, S.; Crocker, J.G.

    1981-01-01

    Present trends in magnetic fusion research and development indicate the promise of commercialization of one of a limited number of inexhaustible energy options early in the next century. Operation of the large-scale fusion experiments, such as the Joint European Torus (JET) and Takamak Fusion Test Reactor (TFTR) now under construction, are expected to achieve the scientific break even point. Early design concepts of power producing reactors have provided problem definition, whereas the latest concepts, such as STARFIRE, provide a desirable set of answers for commercialization. Safety and environmental concerns have been considered early in the development of magnetic fusion reactor concepts and recognition of proplem areas, coupled with a program to solve these problems, is expected to provide the basis for safe and environmentally acceptable commercial reactors. First generation reactors addressed in this paper are expected to burn deuterium and tritium fuel because of the relatively high reaction rates at lower temperatures compared to advanced fuels such as deuterium-deuterium. This paper presents an overwiew of the safety and environmental problems presently perceived, together with some of the programs and techniques planned and/or underway to solve these problems. A preliminary risk assessment of fusion technology relative to other energy technologies is made. Improvements based on material selection are discussed. Tritium and neutron activation products representing potential radiological hazards in fusion reactor are discussed, and energy sources that can lead to the release of radioactivity from fusion reactors under accident conditions are examined. The handling and disposal of radioactive waste are discussed; the status of biological effects of magnetic fields are referenced; and release mechanisms for tritium and activation products, including analytical methods, are presented. (orig./GG)

  19. Sextant: an expert system for transient analysis of nuclear reactors and integral test facilities

    International Nuclear Information System (INIS)

    Barbet, N.; Dumas, M.; Mihelich, G.

    1987-01-01

    Expert systems provide a new way of dealing with the computer-aided management of nuclear plants by combining several knowledge bases and reasoning modes together with a set of numerical models for real-time analysis of transients. New development tools are required together with metaknowledge bases handling temporal hypothetical reasoning and planning. They have to be efficient and robust because during a transient, neither measurements nor models, nor scenarios are hold as absolute references. SEXTANT is a general purpose physical analyzer intended to provide a pattern and avoid duplication of general tools and knowledge bases for similar applications. It combines several knowledge bases concerning measurements, models and qualitative behavior of PWR with a mechanism of conjecture-refutation and a set of simplified models matching the current physical state. A prototype is under assessment by dealing with integral test facility transients. For its development, SEXTANT requires a powerful shell. SPIRAL is such a toolkit, oriented towards online analysis of complex processes and already used in several applications

  20. LWR fuel rod testing facilities in high flux reactor (HFT) Petten for investigation of power cycling and ramping behaviour

    Energy Technology Data Exchange (ETDEWEB)

    Markgraf, J; Perry, D; Oudaert, J [Commission of the European Communities, Joint Reserach Centre, Petten Establishment, Petten (Netherlands)

    1983-06-01

    LWR fuel rod irradiation experiments are being performed in HFR's Pool Side Facility (PSF) by means of pressurized boiling water capsules (BWFC). For more than 6 years the major application of these devices has been in performing irradiation programs to investigate the power ramp behaviour of PWR and BWR rods which have been pre-irradiated in power reactors. Irradiation devices with various types of monitoring instrumentation are available, e.g. for fuel rod length, fuel stack length, fuel rod internal pressure and fuel rod central temperature measurements. The application scope covers PWR and BWR fuel rods, with burn-up levels up to 45 MWd/kg(U), max. linear heat generation rates up to 700 W/cm and simulation of constant power change rates between 0.05 and 1000 W/cm.min. The paper describes the different designs of LWR fuel rod testing facilities and associated non-destructive testing devices in use at the HFR Petten. It also addresses the new test capabilities that will become available after exchange of the HFR vessel in 1983. Furthermore it shows some typical results. (author)

  1. Analysis on working pressure selection of ACME integral test facility

    International Nuclear Information System (INIS)

    Chen Lian; Chang Huajian; Li Yuquan; Ye Zishen; Qin Benke

    2011-01-01

    An integral effects test facility, advanced core cooling mechanism experiment facility (ACME) was designed to verify the performance of the passive safety system and validate its safety analysis codes of a pressurized water reactor power plant. Three test facilities for AP1000 design were introduced and review was given. The problems resulted from the different working pressures of its test facilities were analyzed. Then a detailed description was presented on the working pressure selection of ACME facility as well as its characteristics. And the approach of establishing desired testing initial condition was discussed. The selected 9.3 MPa working pressure covered almost all important passive safety system enables the ACME to simulate the LOCAs with the same pressure and property similitude as the prototype. It's expected that the ACME design would be an advanced core cooling integral test facility design. (authors)

  2. Probabilistic safety assessment for research reactors

    International Nuclear Information System (INIS)

    1986-12-01

    Increasing interest in using Probabilistic Safety Assessment (PSA) methods for research reactor safety is being observed in many countries throughout the world. This is mainly because of the great ability of this approach in achieving safe and reliable operation of research reactors. There is also a need to assist developing countries to apply Probabilistic Safety Assessment to existing nuclear facilities which are simpler and therefore less complicated to analyse than a large Nuclear Power Plant. It may be important, therefore, to develop PSA for research reactors. This might also help to better understand the safety characteristics of the reactor and to base any backfitting on a cost-benefit analysis which would ensure that only necessary changes are made. This document touches on all the key aspects of PSA but placed greater emphasis on so-called systems analysis aspects rather than the in-plant or ex-plant consequences

  3. Instrumentation and measurement method for the ATLAS test facility

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Byong Jo; Chu, In Chul; Eu, Dong Jin; Kang, Kyong Ho; Kim, Yeon Sik; Song, Chul Hwa; Baek, Won Pil

    2007-03-15

    An integral effect test loop for pressurized water reactors (PWRs), the ATLAS is constructed by thermal-hydraulic safety research division in KAERI. The ATLAS facility has been designed to have the length scale of 1/2 and area scale of 1/144 compared with the reference plant, APR1400 which is a Korean evolution type nuclear reactors. A total 1300 instrumentations is equipped in the ATLAS test facility. In this report, the instrumentation of ATLAS test facility and related measurement methods were introduced.

  4. Progress in design, research and development and testing of safety systems for advanced water cooled reactors. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1996-04-01

    The meeting covered the following topics: Developments in design of safety-related heat removal components and systems for advanced water cooled reactors; status of test programmes on heat removal components and systems of new designs; range of validity and extrapolation of test results for the qualification of design/licensing computer models and codes for advanced water cooled reactors; future needs and trends in testing of safety systems for advanced water cooled reactors. Tests of heat removal safety systems have been conducted by various groups supporting the design, testing and certification of advanced water cooled reactors. The Technical Committee concluded that the reported test results generally confirm the predicted performance features of the advanced designs. Refs, figs, tabs

  5. Progress in design, research and development and testing of safety systems for advanced water cooled reactors. Proceedings of a technical committee meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-04-01

    The meeting covered the following topics: Developments in design of safety-related heat removal components and systems for advanced water cooled reactors; status of test programmes on heat removal components and systems of new designs; range of validity and extrapolation of test results for the qualification of design/licensing computer models and codes for advanced water cooled reactors; future needs and trends in testing of safety systems for advanced water cooled reactors. Tests of heat removal safety systems have been conducted by various groups supporting the design, testing and certification of advanced water cooled reactors. The Technical Committee concluded that the reported test results generally confirm the predicted performance features of the advanced designs. Refs, figs, tabs.

  6. Surveys of research projects concerning nuclear facility safety, financed by the Federal Ministry for the Environment, Nature Protection and Reactor Safety, 1988

    International Nuclear Information System (INIS)

    1989-11-01

    Each progress report is a collection of individual reports, categorized by subject matter. They are a documentation of the contractor's progress, rendered by themselves on standardized forms, published, for the sake of general information on progress made in investigations concerning reactor safety, by the project attendance department of the GRS. The individual reports have serial numbers. Each report includes particulars of the objective, work carried out, results obtained and plans for project continuation. (orig.) [de

  7. Surveys of research projects concerning nuclear facility safety, financed by the Federal Ministry for the Environment, Nature Protection and Reactor Safety, 1987

    International Nuclear Information System (INIS)

    1988-06-01

    Each progress report is a collection of individual reports, categorized by subject matter. They are a documentation of the contractor's progress, rendered by themselves on standardized forms, published, for the sake of general information on progress made in investigations concerning reactor safety, by the project attendance department of the GRS. The individual reports have serial numbers. Each report includes particulars of the objective, work carried out, results obtained and plans for project continuation. (orig.) [de

  8. Surveys of research projects concerning nuclear facility safety financed by the Federal Ministry for the Environment, Nature Protection and Reactor Safety, 1991

    International Nuclear Information System (INIS)

    1992-09-01

    Each progress report is a collection of individual reports, categorized by subject matter. They are a documentation of the contractor's progress, rendered by themselves on standardized forms, published, for the sake of general information on progress made in investigations concerning reactor safety, by the project attendance department of the GRS. The individual reports have serial numbers. Each report includes particulars of the objective, work carried out, results obtained and plans for project continuation. (orig.) [de

  9. Development of a test facility for analyzing transients in supercritical water-cooled reactors by fractional scaling analysis

    Energy Technology Data Exchange (ETDEWEB)

    Roberto, Thiago D., E-mail: thiagodbtr@gmail.com [Instituto de Engenharia Nuclear (IEN/CNEN—RJ), Rua Hélio de Almeida, 75 21941-972, Rio de Janeiro Caixa-Postal: 68550, RJ (Brazil); Silva, Mário A. B. da, E-mail: mabs500@gmail.com [Departamento de Energia Nuclear (CTG/UFPE), Av. Professor Luiz Freire, 1000, Recife 50740-540, PE (Brazil); Lapa, Celso M.F., E-mail: lapa@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN—RJ), Rua Hélio de Almeida, 75 21941-972, Rio de Janeiro Caixa-Postal: 68550, RJ (Brazil)

    2016-01-15

    The feasibility of performing experiments using water under supercritical conditions is limited by technical and financial difficulties. These difficulties can be overcome by using model fluids that are characterized by feasible supercritical conditions, that is, lower critical pressure and critical temperature. Experimental investigations are normally used to determine the conditions under which model fluids reliably represent supercritical fluids under steady-state conditions. A fluid-to-fluid scaling approach has been proposed to determine the model fluids that represent supercritical fluids in a transient state. Recently, a similar technique known as fractional scaling analysis was developed to establish the conditions under which experiments can be performed using models that represent transients in prototypes. This paper presents a fractional scaling analysis application to determine parameters for a test facility in which transient conditions in supercritical water-cooled reactors are simulated by using carbon dioxide as a model fluid, whose critical point conditions are more feasible than those of water. Similarity is obtained between water (prototype) and carbon dioxide (model) by depressurization in a simple vessel. The main parameters required for the construction of a future test facility are obtained using the proposed method.

  10. Development of a test facility for analyzing transients in supercritical water-cooled reactors by fractional scaling analysis

    International Nuclear Information System (INIS)

    Roberto, Thiago D.; Silva, Mário A. B. da; Lapa, Celso M.F.

    2016-01-01

    The feasibility of performing experiments using water under supercritical conditions is limited by technical and financial difficulties. These difficulties can be overcome by using model fluids that are characterized by feasible supercritical conditions, that is, lower critical pressure and critical temperature. Experimental investigations are normally used to determine the conditions under which model fluids reliably represent supercritical fluids under steady-state conditions. A fluid-to-fluid scaling approach has been proposed to determine the model fluids that represent supercritical fluids in a transient state. Recently, a similar technique known as fractional scaling analysis was developed to establish the conditions under which experiments can be performed using models that represent transients in prototypes. This paper presents a fractional scaling analysis application to determine parameters for a test facility in which transient conditions in supercritical water-cooled reactors are simulated by using carbon dioxide as a model fluid, whose critical point conditions are more feasible than those of water. Similarity is obtained between water (prototype) and carbon dioxide (model) by depressurization in a simple vessel. The main parameters required for the construction of a future test facility are obtained using the proposed method.

  11. Monitor for safety engineering facility

    International Nuclear Information System (INIS)

    Sato, Akira; Kaneda, Mitsunori.

    1982-01-01

    Purpose: To improve the reactor safety and decrease misoperation upon periodical inspection by instantly obtaining the judgement for the stand-by states in engineering safety facilities of a nuclear power plant. Constitution: Process inputs representing the states of valves, pumps, flowrates or the likes of the facility are gathered into an input device and inputted to a status monitor. The status of the facility inputted to the input device are judged for each of the inputs in a judging section and recognized as a present system stand-by pattern of the system (Valve) to be inspected. While on the other hand, a normal system stand-by pattern previously stored in a memory unit is read out by an instruction from an operator console and judged by comparison with the system stand-by pattern in a comparison section. The results are displayed on a display device. Upon periodical inspection, inspection procedures stored in the memory unit are displayed on the display device by the instruction from the operator console. (Seki, T.)

  12. Operational experience on sodium deposits in KNK reactor and RSB test facility

    Energy Technology Data Exchange (ETDEWEB)

    Jansing, W; Kirchner, G; Menck, J [INTERATOM, Bergisch Gladbach (Germany)

    1977-01-01

    A specific problem of sodium-cooled reactor plants is the formation of sodium aerosols which deposit at cold sections of the plant. Formation and behaviour of sodium aerosols depend on various factors. These may show extreme different effects under conditions which first seem to be identical. Thus, it is very difficult to set up general valid rules on sodium aerosols. By operational experience gained in different plants under divers operating conditions, knowledge is drawn which corresponds well with theoretical considerations. (author)

  13. Investigation of coolant mixture in pressurized water reactors at the Rossendorf mixing test facility ROCOM

    International Nuclear Information System (INIS)

    Grunwald, G.; Hoehne, T.; Prasser, H.M.; Richter, K.; Weiss, F.P.

    1999-01-01

    During the so-called boron dilution or cold water transients at pressurized water reactors too weakly borated water or too cold water, respectively, might enter the reactor core. This results in the insertion of positive reactivity and possibly leads to a power excursion. If the source of unborated or subcooled water is not located in all coolant loops but in selected ones only, the amount of reactivity insertion depends on the coolant mixing in the downcomer and lower plenum of the reactor pressure vessel (RPV). Such asymmetric disturbances of the coolant temperature or boron concentration might e.g. be the result of a failure of the chemical and volume control system (CVCS) or of a main steam line break (MSLB) that does only affect selected steam generators (SG). For the analysis of boron dilution or MSLB accidents coupled neutron kinetics/thermo-hydraulic system codes have been used. To take into account coolant mixing phenomena in these codes in a realistic manner, analytical mixing models might be included. These models must be simple and fast running on the one hand, but must well describe the real mixing conditions on the other hand. (orig.)

  14. RB research reactor safety report

    International Nuclear Information System (INIS)

    Sotic, O.; Pesic, M.; Vranic, S.

    1979-04-01

    This new version of the safety report is a revision of the safety report written in 1962 when the RB reactor started operation after reconstruction. The new safety report was needed because reactor systems and components have been improved and the administrative procedures were changed. the most important improvements and changes were concerned with the use of highly enriched fuel (80% enriched), construction of reactor converter outside the reactor vessel, improved control system by two measuring start-up channels, construction of system for heavy water leak detection, new inter phone connection between control room and other reactor rooms. This report includes description of reactor building with installations, rector vessel, reactor core, heavy water system, control system, safety system, dosimetry and alarm systems, experimental channels, neutron converter, reactor operation. Safety aspects contain analyses of accident reasons, method for preventing reactivity insertions, analyses of maximum hypothetical accidents for cores with natural uranium, 2% enriched and 80% enriched fuel elements. Influence of seismic events on the reactor safety and well as coupling between reactor and the converter are parts of this document

  15. Description of the blowdown test facility COG program on in-reactor fission product release, transport, and deposition under severe accident conditions

    International Nuclear Information System (INIS)

    Fehrenbach, P.J.; Wood, J.C.

    1987-06-01

    Loss-of-coolant accidents with additional impairment of emergency cooling would probably result in high fuel temperatures leading to severe fuel damage (SFD) and significant fission product activity would then be transported along the PHTS to the break where a fraction of it would be released and transport under such conditions, there are many interacting and sometimes competing phenomena to consider. Laboratory simulations are being used to provide data on these individual phenomena, such as UO 2 oxidation and Zr-UO 2 interaction, from which mathematical models can be constructed. These are then combined into computer codes to include the interaction effects and assess the overall releases. In addition, in-reactor tests are the only source of data on release and transport of short-lived fission product nuclides, which are important in the consequence analysis of CANDU reactor accidents. Post-test decontamination of an in-reactor test facility also provides a unique opportunity to demonstrate techniques and obtain decontamination data relevant to post-accident rehabilitation of CANDU power reactors. Specialized facilities are required for in-reactor testing because of the extensive release of radioactive fission products and the high temperatures involved (up to 2500 degrees Celsius). To meet this need for the Canadian program, the Blowdown Test Facility (BTF) has been built in the NRU reactor at Chalk River. Between completion of construction in mid-1987 and the first Zircaloy-sheathed fuel test in fiscal year 1987/88, several commissioning tests are being performed. Similarly, extensive development work has been completed to permit application of instrumentation to irradiated fuel elements, and in support of post-test fuel assembly examination. A program of decontamination studies has also been developed to generate information relevant to post-accident decontamination of power reactors. The BTF shared cost test program funded by the COG High Temperature

  16. Safety status of Russian research reactors

    International Nuclear Information System (INIS)

    Morozov, S.I.

    2001-01-01

    Gosatomnadzor of Russia is conducting the safety regulation and inspection activity related to nuclear and radiation safety at nuclear research facilities, including research reactors, critical assemblies and sub-critical assemblies. It implies implementing three major activities: 1) establishing the laws and safety standards in the field of research reactors nuclear and radiation safety; 2) research reactors licensing; and 3) inspections (or license conditions tracking and inspection). The database on nuclear research facilities has recently been updated based on the actual status of all facilities. It turned out that many facilities have been shutdown, whether temporary or permanently, waiting for the final decision on their decommissioning. Compared to previous years the situation has been inevitably changing. Now we have 99 nuclear research facilities in total under Gosatomnadzor of Russia supervision (compared to 113 in previous years). Their distribution by types and operating organizations is presented. The licensing and conduct of inspection processes are briefly outlined with emphasis being made on specific issues related to major incidents that happened in 2000, spent fuel management, occupational exposure, effluents and emissions, emergency preparedness and physical protection. Finally, a summary of problems at current Russian research facilities is outlined. (author)

  17. Buildings, fields of activity, testing facilities

    International Nuclear Information System (INIS)

    1974-01-01

    Since 1969 the activities of the Materialpruefungsanstalt Stuttgart (MPA) have grown quickly as planned, especially in the field of reactor safety research, which made it necessary to increase the staff to approximately 165 members, to supplement the machines and equipment and to extend the fields of activities occasioning a further departmental reorganization. At present the MPA has the following departments: 1. Teaching (materials testing, materials science and strength of materials) 2. Materials and Welding Technology 3. Materials Science and General Materials Testing with Tribology 4. Design and Strength 5. Creep and Fatigue Testing 6. Central Facilities 7. Vessel and Component Testing. (orig./RW) [de

  18. Thermal hydraulic test for reactor safety system; a visualization study on flow boiling and bubble behavior

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Soon Heung; Baek, Won Pil; Ban, In Cheol [Korea Advanced Institute of Science and Technology, Taejeon (Korea)

    2002-03-01

    The project contribute to understand and to clarify the physical mechanism of flow nucleate boiling and CHF phenomena through the visualization experiments. the results are useful in the development of the enhancement device of heat transfer and to enhance nuclear fuel safety 1. Visual experimental facility 2. Application method of visualization Technique 3. Visualization results of flow nucleate boiling regime - Overall Bubble Behavior on the Heated Surface - Bubble Behavior near CHF Condition - Identification of Flow Structure - Three-layer flow structure 4. Quantifying of bubble parameter through a digital image processing - Image Processing Techniques - Classification of objects and measurements of the size - Three dimensional surface plot with using the luminance 5. Development and estimation of a correlation between bubble diameter and flow parameter - The effect of system parameter on bubble diameter - The development of a bubble diameter correlation . 49 refs., 42 figs., 7 tabs. (Author)

  19. Outlines and verifications of the codes used in the safety analysis of High Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Shiina, Yasuaki; Kunitomi, Kazuhiko; Maruyama, Soh; Fujita, Shigeki; Nakagawa, Shigeaki; Iyoku, Tatsuo; Shindoh, Masami; Sudo, Yukio; Hirano, Masashi.

    1990-03-01

    This paper presents brief description of the computer codes used in the safety analysis of High Temperature Engineering Test Reactor. The list of the codes is: 1. BLOOST-J2 2. THYDE-HTGR 3. TAC-NC 4. RATSAM6 5. COMPARE-MOD1 6. GRACE 7. OXIDE-3F 8. FLOWNET/TRUMP. Of described above, 1, 3, 4, 5, 6 and 7 were developed for the multi-hole type gas cooled reactor and improved for HTTR and 2 was originated by THYDE-codes which were developed to treat the transient thermo-hydraulics during LOCA of LWR. Each code adopted the models and properties which yield conservative analytical results. Adequacy of each code was verified by the comparison with the experimental results and/or the analytical results obtained from the other codes which were already proven. (author)

  20. EURAC: the JRC proposal for an European fusion reactor materials test and development facility

    International Nuclear Information System (INIS)

    Kley, W.; Bishop, G.R.

    1986-01-01

    For the last 7 years we examined the use of a Spallation Neutron Source (SNS) as an altenative European Option to FMIT. For an optimized spallation neutron source design we find now for the same beam power the following design parameters: - Linear Accelerator: 600 MeV, 6 m-A-proton beam on liquid lead target - irradiation parameters: 320 dpa/year in 20 cm 3 or 274 dpa/year in 31.5 cm 3 6 -1 sec -1 in order to simulate the Pulsed Mode of Tokamak Power Reactors. The deflected beam can be used for other experiments

  1. Reactor water injection facility

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Kazuhiro; Kinoshita, Shoichiro

    1997-05-02

    A steam turbine and an electric generator are connected by way of a speed convertor. The speed convertor is controlled so that the number of rotation of the electric generator is constant irrespective of the speed change of the steam turbine. A shaft coupler is disposed between the turbine and the electric generator or between the turbine and a water injection pump. With such a constitution, the steam turbine and the electric generator are connected by way of the speed convertor, and since the number of revolution of the electric generator is controlled to be constant, the change of the number of rotation of the turbine can be controlled irrespective of the change of the number of rotation of the electric generator. Accordingly, the flow rate of the injection water from the water injection pump to a reactor pressure vessel can be controlled freely thereby enabling to supply stable electric power. (T.M.)

  2. Safety device for nuclear reactor

    International Nuclear Information System (INIS)

    Jacquelin, Roland.

    1977-01-01

    This invention relates to a safety device for a nuclear reactor, particularly a liquid metal (generally sodium) cooled fast reactor. This safety device includes an absorbing element with a support head connected by a disconnectable connector formed by the armature of an electromagnet at the end of an axially mobile vertical control rod. This connection is so designed that in the event of it becoming disconnected, the absorbing element gravity slides in a passage through the reactor core into an open container [fr

  3. Safety research experiment facilities, Idaho National Engineering Laboratory, Idaho. Final environmental impact statement

    International Nuclear Information System (INIS)

    Liverman, J.L.

    1977-09-01

    This environmental statement was prepared for the Safety Research Experiment Facilities (SAREF) Project. The purpose of the proposed project is to modify some existing facilities and provide a new test facility at the Idaho National Engineering Laboratory (INEL) for conducting fast breeder reactor (FBR) safety experiments. The SAREF Project proposal has been developed after an extensive study which identified the FBR safety research needs requiring in-reactor experiments and which evaluated the capability of various existing and new facilities to meet these needs. The proposed facilities provide for the in-reactor testing of large bundles of prototypical FBR fuel elements under a wide variety of conditions, ranging from those abnormal operating conditions which might be expected to occur during the life of an FBR power plant to the extremely low probability, hypothetical accidents used in the evaluation of some design options and in the assessment of the long-term potential risk associated with wide-acale deployment of the FBR

  4. Nuclear reactor safety. Quarterly progress report, October 1--December 31, 1977

    International Nuclear Information System (INIS)

    Jackson, J.F.; Stevenson, M.G.

    1978-02-01

    Progress in reactor safety research is summarized. LWR studies include TRAC code development for thermal-hydraulic analysis of accidents, containment systems evaluation, and safety experiments. LMFBR studies include SIMMER code development and applications, modeling of core disruptive accidents, and safety test facilities studies. HTGR safety studies cover fission product release and transport, structural evaluation, phenomena modeling, systems analysis, and accident delineation. GCFR studies are focussed on core disruptive testing

  5. Emergency reactor core cooling facility

    International Nuclear Information System (INIS)

    Yoshikawa, Kazuhiro; Kinoshita, Shoichiro; Iwata, Yasutaka.

    1996-01-01

    The present invention provides an emergency reactor core cooling device for a BWR type nuclear power plant. Namely, D/S pit (gas/water separator storage pool) water is used as a water source for the emergency reactor core cooling facility upon occurrence of loss of coolant accidents (LOCA) by introducing the D/S pit water to the emergency reactor core cooling (ECCS) pump. As a result, the function as the ECCS facility can be eliminated from the function of the condensate storage tank which has been used as the ECCS facility. If the function is unnecessary, the level of quality control and that of earthquake resistance of the condensate storage tank can be lowered to a level of ordinary facilities to provide an effect of reducing the cost. On the other hand, since the D/S pit as the alternative water source is usually a facility at high quality control level and earthquake resistant level, there is no problem. The quality of the water in the D/S pit can be maintained constant by elevating pressure of the D/S pit water by a suppression pool cleanup (SPCU) pump to pass it through a filtration desalter thereby purifying the D/S pit water during the plant operation. (I.S.)

  6. Emergency reactor core cooling facility

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Kazuhiro; Kinoshita, Shoichiro; Iwata, Yasutaka

    1996-11-01

    The present invention provides an emergency reactor core cooling device for a BWR type nuclear power plant. Namely, D/S pit (gas/water separator storage pool) water is used as a water source for the emergency reactor core cooling facility upon occurrence of loss of coolant accidents (LOCA) by introducing the D/S pit water to the emergency reactor core cooling (ECCS) pump. As a result, the function as the ECCS facility can be eliminated from the function of the condensate storage tank which has been used as the ECCS facility. If the function is unnecessary, the level of quality control and that of earthquake resistance of the condensate storage tank can be lowered to a level of ordinary facilities to provide an effect of reducing the cost. On the other hand, since the D/S pit as the alternative water source is usually a facility at high quality control level and earthquake resistant level, there is no problem. The quality of the water in the D/S pit can be maintained constant by elevating pressure of the D/S pit water by a suppression pool cleanup (SPCU) pump to pass it through a filtration desalter thereby purifying the D/S pit water during the plant operation. (I.S.)

  7. Safety upgrades to the NRU research reactor

    International Nuclear Information System (INIS)

    DeAbreu, B.; Mark, J.M.; Mutterback, E.J.

    1998-01-01

    The NRU (National Research Universal) Reactor is a 135 MW thermal research facility located at Chalk River Laboratories, and is owned and operated by Atomic Energy of Canada Limited. One of the largest and most versatile research reactors in the world, it serves as the R and D workhorse for Canada's CANDU business while at the same time filling the role as one of the world's major producers of medical radioisotopes. AECL plans to extend operation of the NRU reactor to approximately the year 2005 when a new replacement, the Irradiation Research Facility (IRF) will be available. To achieve this, AECL has undertaken a program of safety reassessment and upgrades to enhance the level of safety consistent with modem requirements. An engineering assessment/inspection of critical systems, equipment and components was completed and seven major safety upgrades are being designed and installed. These upgrades will significantly reduce the reactor's vulnerability to common mode failures and external hazards, with particular emphasis on seismic protection. The scheduled completion date for the project is 1999 December at a cost approximately twice the annual operating cost. All work on the NRU upgrade project is planned and integrated into the regular operating cycles of the reactor; no major outages are anticipated. This paper describes the safety upgrades and discusses the technical and managerial challenges involved in extending the operating life of the NRU reactor. (author)

  8. Reactor safety in Eastern Europe

    International Nuclear Information System (INIS)

    1995-02-01

    The papers presented to the GRS colloquium refer to the cooperative activities for reactor accident analysis and modification of the GRS computer codes for their application to reactors of the Russian design types of WWER or RBMK. Another topic is the safety of RBMK reactors in particular, and the current status of investigations and studies addressing the containment of unit 4 of the Chernobyl reactor station. All papers are indexed separately in report GRS--117. (HP)

  9. Real time simulator for material testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Takemoto, Noriyuki; Imaizumi, Tomomi; Izumo, Hironobu; Hori, Naohiko; Suzuki, Masahide [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan); Ishitsuka, Tatsuo; Tamura, Kazuo [ITOCHU Techno-Solutions Corp., Tokyo (Japan)

    2012-03-15

    Japan Atomic Energy Agency (JAEA) is now developing a real time simulator for a material testing reactor based on Japan Materials Testing Reactor (JMTR). The simulator treats reactor core system, primary and secondary cooling system, electricity system and irradiation facility systems. Possible simulations are normal reactor operation, unusual transient operation and accidental operation. The developed simulator also contains tool to revise/add facility in it for the future development. (author)

  10. Real time simulator for material testing reactor

    International Nuclear Information System (INIS)

    Takemoto, Noriyuki; Imaizumi, Tomomi; Izumo, Hironobu; Hori, Naohiko; Suzuki, Masahide; Ishitsuka, Tatsuo; Tamura, Kazuo

    2012-01-01

    Japan Atomic Energy Agency (JAEA) is now developing a real time simulator for a material testing reactor based on Japan Materials Testing Reactor (JMTR). The simulator treats reactor core system, primary and secondary cooling system, electricity system and irradiation facility systems. Possible simulations are normal reactor operation, unusual transient operation and accidental operation. The developed simulator also contains tool to revise/add facility in it for the future development. (author)

  11. Reactor fuel exchanging facility

    International Nuclear Information System (INIS)

    Kubota, Shin-ichi.

    1981-01-01

    Purpose: To enable operation of an emergency manual operating mechanism for a fuel exchanger with all operatorless trucks and remote operation of a manipulator even if the exchanger fails during the fuel exchanging operation. Constitution: When a fuel exchanging system fails while connected to a pressure tube of a nuclear reactor during a fuel exchanging operation, a stand-by self-travelling truck automatically runs along a guide line to the position corresponding to the stopping position at that time of the fuel exchanger based on a command from a central control chamber. At this time the truck is switched to manual operation, and approaches the exchanger while being monitored through a television camera and then stops. Then, a manipurator is connected to the emergency manual operating mechanism of the exchanger, and is operated through necessary emergency steps by driving the snout, the magazine, the grab or the like in the exchanger in response to the problem, and necessary operations for the emergency treatment are thus performed. (Sekiya, K.)

  12. Emergency reactor container cooling facility

    International Nuclear Information System (INIS)

    Suzuki, Hiroaki; Matsumoto, Tomoyuki.

    1992-01-01

    The present invention concerns an emergency cooling facility for a nuclear reactor container having a pressure suppression chamber, in which water in the suppression chamber is effectively used for cooling the reactor container. That is, the lower portion of a water pool in the pressure suppression chamber and the inside of the reactor container are connected by a pipeline. The lower end of the pipeline and a pressurized incombustible gas tank disposed to the outside of the reactor container are connected by a pipeline by way of valves. Then, when the temperature of the lower end of the pressure vessel exceeds a predetermined value, the valves are opened. If the valves are opened, the incombustible gas flows into the lower end of the pipeline connecting the lower portion of the water pool in the pressure suppression chamber and the inside of the reactor container. Since the inside of the pipeline is a two phase flow comprising a mixture of a gas phase and a liquid phase, the average density is decreased. Therefore, the water level of the two phase flow is risen by the level difference between the inside and the outside of the pipeline and, finally, the two phase mixture is released into the reactor container. As a result, the reactor container can be cooled by water in the suppression chamber by a static means without requiring pumps. (I.S.)

  13. Radiation protection planning for decommissioning of research reactor facilities

    International Nuclear Information System (INIS)

    Jackson, Roger; Harman, Neil; Craig, David; Fecitt, Lorna; Lobach, Yuri; Gorlinskij, Juri; Kolyadin, Vyacheslav; Pavlenko, Vytali

    2008-01-01

    The MR reactor at the Russian Research Centre Kurchatov Institute (RRCKI), Moscow was a 50 MW multipurpose material testing and research reactor equipped with nine experimental loop facilities to test prototype fuel for various nuclear power reactors being developed. The reactor was shut down in 1993 and de-fuelled. The experimental loops are located in basement rooms around the reactor. The nature of the research into the characteristics of fuel design and coolant chemistry resulted in fission products and activation products in the test loop equipment. Decommissioning of the loops therefore presents a number of challenges. In addition the city of Moscow has expanded such that the RRC KI is now surrounded by housing which had to be taken into account in the radiological protection planning. This paper describes the techniques proposed to undertake the dismantling operations in order to minimise the radiation exposure to workers and members of the public. Estimates have been made of the worker doses which could be incurred during the dismantling process and the environmental impacts which could occur. These are demonstrated to be as low as reasonably achievable. The work was funded by the UK Department of Business Enterprise and Regulatory Reform (DBERR) (formerly the Department of Trade and Industry) under the Nuclear Safety Programme (NSP) set up to address nuclear safety issues in the Former Soviet Union. (author)

  14. Safety in connection with the request for approval of the installation alteration in the fuel reprocessing facilities of Power Reactor and Nuclear Fuel Development Corporation (report)

    International Nuclear Information System (INIS)

    1982-01-01

    A report to the Prime Minister by the Nuclear Safety Commission was presented concerning the safety in the installation alteration of the fuel reprocessing facilities, as PNC had requested its approval to the Prime Minister. The safety was confirmed. The items of examination on the safety made by the committee on Examination of Nuclear Fuel Safety of NSC were the aseismic design of liquid waste storage, uranium denitration facility, intermediate gate and radioactive solid waste storage; the criticality safety design of the denitration facility; the radiation shielding design of the liquid waste storage, denitration facility and solid waste storage; the function of radioactive material containment of the liquid waste storage and denitration facility; the radiation control in the liquid waste storage, denitration facility and solid waste storage; the waste management in the liquid waste storage and denitration facility; fire and explosion prevention in the liquid waste storage; exposure dose from the liquid waste storage and denitration facility. (Mori, K.)

  15. Nuclear Reactor Facility

    International Nuclear Information System (INIS)

    Schabert, H.P.; Ropers, J.

    1976-01-01

    A pressurized-water reactor pressure vessel connects via a main coolant pipe loop including a main coolant pump, with the lower portion of at least one vertical steam generator horizontally offset from the pressure vessel. This equipment is contained by a concrete structure entirely enclosing the pressure vessel and forming a generator room horizontally enclosing the generator and the loop and extending upwardly to an open top closed by a horizontal ceiling. The concrete structure is completely surrounded by a spherical steel containment shell designed to withstand any internal fluid pressure which might result from an accidental release of the coolant inside of this shell, and the shell forms a large space above the entire concrete structure. The ceiling above the generator room is a horizontal steel gridlike construction defining a plurality of vertical openings which are normally closed by horizontal sheet metal plates which are hinged to the gridlike construction and are light enough in weight to be forced upwardly, to open the openings, when the plates receive upward force from fluid pressure below them resulting from the loop, or other equipment in the generator room, accidentally permitting a sudden release of the pressurized-water coolant. The high fluid pressure that would otherwise develop within the concrete generator room, is in this way almost immediately relieved via the openings of the grid-like construction, by the plates being forced upwardly, the pressure being then dissipated upwardly in the large space above the top of the concrete structure, provided by the steel containment shell. This prevents the upstanding wall portions of the generator room from being stressed, and possibly damaged, by any sudden release of coolant in the generator room. Other features are disclosed

  16. Potential role of the Fast Flux Test Facility and the advanced test reactor in the U.S. tritium production system

    Energy Technology Data Exchange (ETDEWEB)

    Dautel, W.A.

    1996-10-01

    The Deparunent of Energy is currently engaged in a dual-track strategy to develop an accelerator and a conunercial light water reactor (CLWR) as potential sources of tritium supply. New analysis of the production capabilities of the Fast Flux Test Facility (FFTF) at the Hanford Site argues for considering its inclusion in the tritium supply,system. The use of the FFTF (alone or together with the Advanced Test Reactor [ATR] at the Idaho National Engineering Laboratory) as an integral part of,a tritium production system would help (1) ensure supply by 2005, (2) provide additional time to resolve institutional and technical issues associated with the- dual-track strategy, and (3) reduce discounted total life-cycle`costs and near-tenn annual expenditures for accelerator-based systems. The FFRF would also provide a way to get an early start.on dispositioning surplus weapons-usable plutonium as well as provide a source of medical isotopes. Challenges Associated With the Dual-Track Strategy The Departinent`s purchase of either a commercial reactor or reactor irradiation services faces challenging institutional issues associated with converting civilian reactors to defense uses. In addition, while the technical capabilities of the individual components of the accelerator have been proven, the entire system needs to be demonstrated and scaled upward to ensure that the components work toge ther 1548 as a complete production system. These challenges create uncertainty over the ability of the du2a-track strategy to provide an assured tritium supply source by 2005. Because the earliest the accelerator could come on line is 2007, it would have to operate at maximum capacity for the first few years to regenerate the reserves lost through radioactive decay aftei 2005.

  17. Potential role of the Fast Flux Test Facility and the advanced test reactor in the U.S. tritium production system

    International Nuclear Information System (INIS)

    Dautel, W.A.

    1996-01-01

    The Department of Energy is currently engaged in a dual-track strategy to develop an accelerator and a commercial light water reactor (CLWR) as potential sources of tritium supply. New analysis of the production capabilities of the Fast Flux Test Facility (FFTF) at the Hanford Site argues for considering its inclusion in the tritium supply,system. The use of the FFTF (alone or together with the Advanced Test Reactor [ATR] at the Idaho National Engineering Laboratory) as an integral part of,a tritium production system would help (1) ensure supply by 2005, (2) provide additional time to resolve institutional and technical issues associated with the- dual-track strategy, and (3) reduce discounted total life-cycle'costs and near-tenn annual expenditures for accelerator-based systems. The FFRF would also provide a way to get an early start.on dispositioning surplus weapons-usable plutonium as well as provide a source of medical isotopes. Challenges Associated With the Dual-Track Strategy The Department's purchase of either a commercial reactor or reactor irradiation services faces challenging institutional issues associated with converting civilian reactors to defense uses. In addition, while the technical capabilities of the individual components of the accelerator have been proven, the entire system needs to be demonstrated and scaled upward to ensure that the components work together 1548 as a complete production system. These challenges create uncertainty over the ability of the du2a-track strategy to provide an assured tritium supply source by 2005. Because the earliest the accelerator could come on line is 2007, it would have to operate at maximum capacity for the first few years to regenerate the reserves lost through radioactive decay after 2005

  18. Tritium Systems Test Facility

    International Nuclear Information System (INIS)

    Cafasso, F.A.; Maroni, V.A.; Smith, W.H.; Wilkes, W.R.; Wittenberg, L.J.

    1978-01-01

    This TSTF proposal has two principal objectives. The first objective is to provide by mid-FY 1981 a demonstration of the fuel cycle and tritium containment systems which could be used in a Tokamak Experimental Power Reactor for operation in the mid-1980's. The second objective is to provide a capability for further optimization of tritium fuel cycle and environmental control systems beyond that which is required for the EPR. The scale and flow rates in TSTF are close to those which have been projected for a prototype experimental power reactor (PEPR/ITR) and will permit reliable extrapolation to the conditions found in an EPR. The fuel concentrations will be the same as in an EPR. Demonstrations of individual components of the deuterium-tritium fuel cycle and of monitoring, accountability and containment systems and of a maintenance methodology will be achieved at various times in the FY 1979-80 time span. Subsequent to the individual component demonstrations--which will proceed from tests with hydrogen (and/or deuterium) through tracer levels of tritium to full operational concentrations--a complete test and demonstration of the integrated fuel processing and tritium containment facility will be performed. This will occur near the middle of FY 1981. Two options were considered for the TSTF: (1) The modification of an existing building and (2) the construction of a new facility

  19. Nordic studies in reactor safety

    International Nuclear Information System (INIS)

    Pershagen, N.

    1993-01-01

    The Nordic Nuclear Safety Research Programme SIK programme in reactor safety is part of a major joint Nordic research effort in nuclear safety. The report summarizes the achievements of the SIK programme, which was carried out during 1990-1993 in collaboration between Nordic nuclear utilities, safety authorities, and research institutes. Three main projects were successfully completed dealing with: 1) development and application of a living PSA concept for monitoring the risk of core damage, and of safety indicators for early warning of possible safety problems; 2) review and intercomparison of severe accident codes, case studies of potential core melt accidents in nordic reactors, development of chemical models for the MAAP code, and outline of a system for computerized accident management support; 3) compilation of information about design and safety features of neighbouring reactors in Germany, Lithuania and Russia, and for naval reactors and nuclear submarines. The report reviews the state-of-the-art in each subject matter as an introduction to the individual project summaries. The main findings of each project are highlighted. The report also contains an overview of reactor safety research in the Nordic countries and a summary of fundamental reactor safety principles. (au) (69 refs.)

  20. LOFT integral test system final safety analysis report

    International Nuclear Information System (INIS)

    1974-03-01

    Safety analyses are presented for the following LOFT Reactor systems: engineering safety features; support buildings and facilities; instrumentation and controls; electrical systems; and auxiliary systems. (JWR)

  1. Decommissioning the UHTREX Reactor Facility at Los Alamos, New Mexico

    International Nuclear Information System (INIS)

    Salazar, M.; Elder, J.

    1992-08-01

    The Ultra-High Temperature Reactor Experiment (UHTREX) facility was constructed in the late 1960s to advance high-temperature and gas-cooled reactor technology. The 3-MW reactor was graphite moderated and helium cooled and used 93% enriched uranium as its fuel. The reactor was run for approximately one year and was shut down in February 1970. The decommissioning of the facility involved removing the reactor and its associated components. This document details planning for the decommissioning operations which included characterizing the facility, estimating the costs of decommissioning, preparing environmental documentation, establishing a system to track costs and work progress, and preplanning to correct health and safety concerns in the facility. Work to decommission the facility began in 1988 and was completed in September 1990 at a cost of $2.9 million. The facility was released to Department of Energy for other uses in its Los Alamos program

  2. Safety Design Criteria of Indian Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Pillai, P.; Chellapandi, P.; Chetal, S.C.; Vasudeva Rao, P.R.

    2013-01-01

    • Important feedback has been gained through the design and safety review of PFBR. • The safety criteria document prepared by AERB and IGCAR would provide important input to prepare the dedicated document for the Sodium cooled Fast Reactors at the national and international level. • A common approach with regard to safety, among countries pursuing fast reactor program, is desirable. • Sharing knowledge and experimental facilities on collaborative basis. • Evolution of strong safety criteria – fundamental to assure safety

  3. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baek, W. P.; Song, C. H.; Kim, Y. S. and others

    2005-02-15

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8{approx}2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4{approx}2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure.

  4. Gas-cooled fast-breeder reactor. Helium Circulator Test Facility updated design cost estimate

    International Nuclear Information System (INIS)

    1979-04-01

    Costs which are included in the cost estimate are: Titles I, II, and III Architect-Engineering Services; Titles I, II, and III General Atomic Services; site clearing, grading, and excavation; bulk materials and labor of installation; mechanical and electrical equipment with installation; allowance for contractors' overhead, profit, and insurance; escalation on materials and labor; a contingency; and installation of GAC supplied equipment and materials. The total estimated cost of the facility in As Spent Dollars is $27,700,000. Also included is a cost comparison of the updated design and the previous conceptual design. There would be a considerable penalty for the direct-cooled system over the indirect-cooled system due to the excessive cost of the large diameter helium loop piping to an outdoor heat exchanger. The indirect cooled system which utilizes a helium/Dowtherm G heat exchanger and correspondingly smaller and lower pressure piping to its outdoor air cooler proved to be the more economical of the two systems

  5. Steam--water mixing in nuclear reactor safety loss-of-coolant experiments

    International Nuclear Information System (INIS)

    Naff, S.A.; Schwarz, W.F.

    1978-01-01

    Computer models used to predict the response of reactors to hypothesized accidents necessarily incorporate approximating assumptions. To verify the models by comparing predicted and measured responses in test facilities, these assumptions must be confirmed to be realistic. Recent experiments in facilities capable of repeatedly duplicating the transient behavior of a pressurized water reactor undergoing a pipe rupture show that the assumption of complete water-steam mixing during the transient results in the predicted decompression being faster than that observed. Water reactor safety studies currently in progress include programs aimed at the verification of computer models or ''codes'' used to predict reactor system responses to various hypothesized accidents. The approach is to compare code predictions of transients with the actual test transients in experimental facilities. The purpose of this paper is to explain an important instance in which predictions and data are not in complete agreement and to indicate the significance to water reactor safety studies

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

    Science.gov (United States)

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

    1997-06-01

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

  7. Test facility for rewetting experiments at CDTN

    International Nuclear Information System (INIS)

    Rezende, Hugo C.; Mesquita, Amir Z.; Ladeira, Luiz C.D.; Santos, Andre A.C.

    2015-01-01

    One of the most important subjects in nuclear reactor safety analysis is the reactor core rewetting after a Loss-of-Coolant Accident (LOCA) in a Light Water Reactor LWR. Several codes for the prediction of the rewetting evolution are under development based on experimental results. In a Pressurized Water Reactor (PWR) the reflooding phase of a LOCA is when the fuel rods are rewetted from the bottom of the core to its top after having been totally uncovered and dried out. Out-of-pile reflooding experiments performed with electrical heated fuel rod simulators show different quench behavior depending the rods geometry. A test facility for rewetting experiments (ITR - Instalacao de Testes de Remolhamento) has been constructed at the Thermal Hydraulics Laboratory of the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), with the objective of performing investigations on basic phenomena that occur during the reflood phase of a LOCA in a PWR, using tubular and annular test sections. This paper presents the design aspects of the facility, and the current stage of the works. The mechanical aspects of the installation as its instrumentation are described. Two typical tests are presented and results compered with theoretical calculations using computer code. (author)

  8. Test facility for rewetting experiments at CDTN

    Energy Technology Data Exchange (ETDEWEB)

    Rezende, Hugo C.; Mesquita, Amir Z.; Ladeira, Luiz C.D.; Santos, Andre A.C., E-mail: hcr@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (SETRE/CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores

    2015-07-01

    One of the most important subjects in nuclear reactor safety analysis is the reactor core rewetting after a Loss-of-Coolant Accident (LOCA) in a Light Water Reactor LWR. Several codes for the prediction of the rewetting evolution are under development based on experimental results. In a Pressurized Water Reactor (PWR) the reflooding phase of a LOCA is when the fuel rods are rewetted from the bottom of the core to its top after having been totally uncovered and dried out. Out-of-pile reflooding experiments performed with electrical heated fuel rod simulators show different quench behavior depending the rods geometry. A test facility for rewetting experiments (ITR - Instalacao de Testes de Remolhamento) has been constructed at the Thermal Hydraulics Laboratory of the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), with the objective of performing investigations on basic phenomena that occur during the reflood phase of a LOCA in a PWR, using tubular and annular test sections. This paper presents the design aspects of the facility, and the current stage of the works. The mechanical aspects of the installation as its instrumentation are described. Two typical tests are presented and results compered with theoretical calculations using computer code. (author)

  9. Safety report on WWR-S reactor

    International Nuclear Information System (INIS)

    Horyna, J.; Kaisler, L.; Listik, E.

    1981-04-01

    The present Safety Report of the WWR-S reactor summarizes findings obtained during the trial and partially also permanent operation of the reactor after two stages of its reconstruction implemented between 1974 and 1976. Most data are presented necessary for assessing probable risks of possible accident conditions whose consequences pose health hazards to individuals of the population, radiation personnel and the facilities themselves. Attention is devoted to the description of the locality, to components and systems, heat removal from the core, design aspects, the quality of new and old parts of the technological circuits, the systems of protection and control, the emergency core cooling system, the problems of radiation safety, and to the safety analyses of the abnormal states envisaged. The Report was compiled with regard to IAEA and CMEA recommendations concerning safe operation of research reactors and to the recommendations and binding decisions of the Czechoslovak Atomic Energy Commission. (author)

  10. A graded approach to safety documentation at processing facilities

    International Nuclear Information System (INIS)

    Cowen, M.L.

    1992-01-01

    Westinghouse Savannah River Company (WSRC) has over 40 major Safety Analysis Reports (SARs) in preparation for non-reactor facilities. These facilities include nuclear material production facilities, waste management facilities, support laboratories and environmental remediation facilities. The SARs for these various projects encompass hazard levels from High to Low, and mission times from startup, through operation, to shutdown. All of these efforts are competing for scarce resources, and therefore some mechanism is required for balancing the documentation requirements. Three of the key variables useful for the decision making process are Depth of Safety Analysis, Urgency of Safety Analysis, and Resource Availability. This report discusses safety documentation at processing facilities

  11. The Advanced Test Reactor Strategic Evaluation Program

    International Nuclear Information System (INIS)

    Buescher, B.J.

    1990-01-01

    A systematic evaluation of safety, environmental, and operational issues has been initiated at the Advanced Test Reactor (ATR). This program, the Strategic Evaluation Program (STEP), provides an integrated review of safety and operational issues against the standards applied to licensed commercial facilities. In the review of safety issues, 18 deviations were identified which required prompt attention. Resolution of these items has been accelerated in the program. An integrated living schedule is being developed to address the remaining findings. A risk evaluation is being performed on the proposed corrective actions and these actions will then be formally ranked in order of priority based on considerations of safety and operational significance. Once the final ranking is completed, an integrated schedule will be developed, which will include considerations of availability of funding and operating schedule. 3 refs., 2 figs

  12. Meeting on reactor safety research

    International Nuclear Information System (INIS)

    1982-09-01

    The meeting 'Reactor Safety Research' organized for the second time by the GRS by order of the BMFT gave a review of research activities on the safety of light water reactors in the Federal Repulbic of Germany, international co-operation in this field and latest results of this research institution. The central fields of interest were subjects of man/machine-interaction, operational reliability accident sequences, and risk. (orig.) [de

  13. Engine Test Facility (ETF)

    Data.gov (United States)

    Federal Laboratory Consortium — The Air Force Arnold Engineering Development Center's Engine Test Facility (ETF) test cells are used for development and evaluation testing of propulsion systems for...

  14. Reactor safety research - results and perspectives

    International Nuclear Information System (INIS)

    Banaschik, M.

    1989-01-01

    The work performed so far is an essential contribution to the determination of the safety margins of nuclear facilities and their systems and to the further development of safety engineering. The further development of safety engineering involves a shift of emphasis in reactor safety research towards event sequences beyond the design basis. The aim of this shift in emphasis is the further development of the preventive level. This is based on the fact that the conservative design of the operating and safety systems involves and essential safety potential. The R and D work is intended to help develop accident management measures and to take the plant back into the safe state even after severe accidents. In this context, it is necessary to make full use of the safety margins of the plant and to include the operating systems for coping with accidents. As a result of the aims, the research work approaches operating and plant-specific processes. (orig./DG) [de

  15. Review of inservice inspection and nondestructive examination practices at DOE Category A test and research reactors

    International Nuclear Information System (INIS)

    Anderson, M.T.; Aldrich, D.A.

    1990-09-01

    In-service inspection (ISI) programs are used at commercial nuclear power plants for monitoring the pressure boundary integrity of various systems and components to ensure their continued safe operation. The Department of Energy (DOE) operates several test and research reactors. This report represents an evaluation of the ISI and nondestructive examination (NDE) practices at five DOE Category A (> 20 MW thermal) reactors as compared, where applicable, to the current ISI activities of commercial nuclear power facilities. The purpose of an inservice inspection (ISI) program is to establish regular surveillance of safety-related components to ensure their safe and reliable operation. The integrity of materials comprising these components is generally monitored by means of periodic nondestructive examinations (NDE), which, if appropriately performed, provide methods for identifying degradation that could render components unable to perform their intended safety functions. The reactors evaluated during this review were the Experimental Breeder Reactor 2 and the Fast Flux Test Facility (liquid-metal cooled plants), the Advanced Test Reactor and the High Flux Isotopes Reactor (light-water cooled reactors), and the High Flux Beam Reactor (a heavy-water cooled facility). Although these facilities are extremely diverse in design and operation, they all have less stored energy, smaller inventories of radionuclides, and generally, more remote locations than commercial reactors. However, all DOE test and research facilities contain components similar to those of commercial reactors for which continued integrity is important to maintain plant safety. 10 refs., 6 tabs

  16. Large-scale testing of in-vessel debris cooling through external flooding of the reactor pressure vessel in the CYBL facility

    International Nuclear Information System (INIS)

    Chu, T.Y.; Bentz, J.H.; Bergeron, K.D.; Slezak, S.E.; Simpson, R.B.

    1994-01-01

    The possibility of achieving in-vessel core retention by flooding the reactor cavity, or the ''flooded cavity'', is an accident management concept currently under consideration for advanced light water reactors (ALWR), as well as for existing light water reactors (LWR). The CYBL (CYlindrical BoiLing) facility is a facility specifically designed to perform large-scale confirmatory testing of the flooded cavity concept. CYBL has a tank-within-a-tank design; the inner 3.7 m diameter tank simulates the reactor vessel, and the outer tank simulates the reactor cavity. The energy deposition on the bottom head is simulated with an array of radiant heaters. The array can deliver a tailored heat flux distribution corresponding to that resulting from core melt convection. The present paper provides a detailed description of the capabilities of the facility, as well as results of recent experiments with heat flux in the range of interest to those required for in-vessel retention in typical ALWRs. The paper concludes with a discussion of other experiments for the flooded cavity applications

  17. Large-Scale testing of in-vessel debris cooling through external flooding of the reactor pressure vessel in the CYBL facility

    International Nuclear Information System (INIS)

    Chu, T.Y.; Bentz, J.H.; Bergeron, K.D.; Slezak, S.E.; Simpson, R.B.

    1994-01-01

    The possibility of achieving in-vessel core retention by flooding the reactor cavity, or the open-quotes flooded cavityclose quotes, is an accident management concept currently under consideration for advanced light water reactors (ALWR), as well as for existing light water reactors (LWR). The CYBL (CYlindrical BoiLing) facility is a facility specifically designed to perform large-scale confirmatory testing of the flooded cavity concept. CYBL has a tank-within-a-tank design; the inner 3.7 m diameter tank simulates the reactor vessel, and the outer tank simulates the reactor cavity. The energy deposition on the bottom head is simulated with an array of radiant heaters. The array can deliver a tailored heat flux distribution corresponding to that resulting from core melt convection. The present paper provides a detailed description of the capabilities of the facility, as well as results of recent experiments with heat flux in the range of interest to those required for in-vessel retention in typical ALWRs. The paper concludes with a discussion of other experiments for the flooded cavity applications

  18. Engineering design of a fusion test reactor (FTR) and fusion engineering research facility (FERF) based on a toroidal theta pinch

    International Nuclear Information System (INIS)

    Abdou, M.; Burke, R.J.; Dauzvardis, P.V.; Foss, M.; Gerstl, S.A.W.; Maroni, V.A.; Pierce, A.W.; Turner, A.F.; Krakowski, R.A.; Linford, R.K.; Oliphant, T.A.; Ribe, F.L.; Thomassen, K.I.

    1975-01-01

    This paper describes two advanced toroidal theta-pinch devices which are being proposed for future construction. The Fusion Test Reactor (FTR) is being designed to produce thermonuclear energy (at 20 MeV/neutron) equal to the maximum plasma energy (Q=1) and to demonstrate α-particle heating. The Fusion Engineering and Research Facility (FERF) is being designed to test materials in a fusion environment where the average 14-MeV neutron flux from the plasma is greater than or of the order of 5.10 13 n/cm 2 .s over large surface areas. These devices employ the staged theta-pinch principle where the heating is accomplished by rapid (about 0.1 μs) implosion and expansion followed by a slow compression of the plasma. The rapid implosion injects as much heat as possible at as large a plasma radious as possible so that the plasma remains stable even after further compression. The final compression to ignition requires the transfer of a large amount of magnetic energy which implies a long transfer time (about 1 ms) for realistic voltages in the driving circuit. Throughout the heating and burn cycle the plasma must remain in equilibrium and stable to the dominant MHD-modes. A sufficiently large plasma radius guarantees stability against the m = 1 modes. These equilibrium and stability conditions and the requirements on thermonuclear burn determine the design parameters for either machine. The design parameters must also be consistent with economic limitations and technological feasibility of components. In addition to these requirements, the FERF must provide a steady and reliable source of fusion neutrons. (author)

  19. Mirror Fusion Test Facility (MFTF)

    International Nuclear Information System (INIS)

    Thomassen, K.I.

    1978-01-01

    A large, new Mirror Fusion Test Facility is under construction at LLL. Begun in FY78 it will be completed at the end of FY78 at a cost of $94.2M. This facility gives the mirror program the flexibility to explore mirror confinement principles at a signficant scale and advances the technology of large reactor-like devices. The role of MFTF in the LLL program is described here

  20. COMPRESS - a computerized reactor safety system

    International Nuclear Information System (INIS)

    Vegh, E.

    1986-01-01

    The computerized reactor safety system, called COMPRESS, provides the following services: scram initiation; safety interlockings; event recording. The paper describes the architecture of the system and deals with reliability problems. A self-testing unit checks permanently the correct operation of the independent decision units. Moreover the decision units are tested by short pulses whether they can initiate a scram. The self-testing is described in detail

  1. Rocketball Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — This test facility offers the capability to emulate and measure guided missile radar cross-section without requiring flight tests of tactical missiles. This facility...

  2. Ten years operating experience at the Fast Flux Test Facility: A decade of excellence

    International Nuclear Information System (INIS)

    Swaim, D.J.; Waldo, J.B.; Farabee, O.A.

    1991-07-01

    The Fast Flux Test Facility is a 400 MW(t) fast reactor cooled by three sodium loops. The Fast Flux Test Facility is managed by the Westinghouse Hanford Company for the US Department of Energy. The Fast Flux Test Facility was designed and constructed to provide irradiation testing of fuels and materials for the US Department of Energy Liquid Metal Reactor research program. Facility activities have increased to include fusion power materials testing, passive safety testing, isotope production, and international collaboration. 5 figs

  3. Concerning safety issue associated with approval of modification of plan for reprocessing facilities of Power Reactor and Nuclear Fuel Development Corporation (reply to inquiry)

    Energy Technology Data Exchange (ETDEWEB)

    1989-01-01

    In response to an inquiry on the title issue received on Sept. 16, 1988, the Nuclear Safety Commission made an adequate study and submitted the findings to the Prime Minister on Oct. 20. the relevant modifications include the installation of combustion facilities comprising two furnaces to replace the existing facilities, and the addition of uranyl nitrate solution receiving equipment to the existing facilities for plutonium conversion technology development. The study addressed the effect of the modifications on the safety of the facilities and the environment. The combustion facilities, equipment and underground seepage water tank are designed in accordance with the applicable anti-earthquake standards. The facilities will be constructed of incombustible of flame-retardant materials. Equipment in the facilities will be used under negative pressure to prevent counterflow. Adequate measures will be taken for shielding, exposure control, waste disposal, etc. It is concluded that the modifications will not affect the safety of the reprocessing facilities. (Nogami, K.).

  4. Concerning safety issue associated with approval of modification of plan for reprocessing facilities of Power Reactor and Nuclear Fuel Development Corporation (reply to inquiry)

    International Nuclear Information System (INIS)

    1989-01-01

    In response to an inquiry on the title issue received on Sept. 16, 1988, the Nuclear Safety Commission made an adequate study and submitted the findings to the Prime Minister on Oct. 20. the relevant modifications include the installation of combustion facilities comprising two furnaces to replace the existing facilities, and the addition of uranyl nitrate solution receiving equipment to the existing facilities for plutonium conversion technology development. The study addressed the effect of the modifications on the safety of the facilities and the environment. The combustion facilities, equipment and underground seepage water tank are designed in accordance with the applicable anti-earthquake standards. The facilities will be constructed of incombustible of flame-retardant materials. Equipment in the facilities will be used under negative pressure to prevent counterflow. Adequate measures will be taken for shielding, exposure control, waste disposal, etc. It is concluded that the modifications will not affect the safety of the reprocessing facilities. (Nogami, K.)

  5. Reactor safety protection system

    International Nuclear Information System (INIS)

    Nishi, Hiroshi; Yokoyama, Tsuguo.

    1989-01-01

    A plurality of neutron detectors are disposed around a reactor core and detection signals from optional two neutron detectors are inputted into a ratio calculation device. If the ratio between both of the neutron flux level signals exceeds a predetermined value, a reactor trip signal is generated from an alarm setting device. Further, detection signals from all of the neutron detection devices are inputted into an average calculation device and the reactor trip signal is generated also in a case where the average value exceeds a predetermined set value. That is, when the reactor core power is increased locally, the detection signal from the neutron detector nearer to the point of power increase is greater than the increase rate for the entire reactor core power, while the detection signal from the neutron detector remote from the point of power increase is smaller. Thus, the local power increase ratio in the FBR reactor core can be detected efficiently by calculating the ratio for the neutron flux level signals from two neutron detectors, thereby enabling to exactly recognize the local power increase rate in the reactor core. (N.H.)

  6. Reactor safety device

    International Nuclear Information System (INIS)

    Okada, Yasumasa.

    1987-01-01

    Purpose: To scram control rods by processing signals from a plurality of temperature detectors and generating abnormal temperature warning upon occurrence of abnormal temperature in a nuclear reactor. Constitution: A temperature sensor comprising a plurality of reactors each having a magnetic body as the magnetic core having a curie point different from each other and corresponding to the abnormal temperature against which reactor core fuels have to be protected is disposed in an identical instrumentation well near the reactor core fuel outlet/inlet of a reactor. A temperature detection device actuated upon detection of an abnormal temperature by the abrupt reduction of the reactance of each of the reactors is disposed. An OR circuit and an AND circuit for conducting OR and AND operations for each of the abnormal temperature detection signals from the temperature detection device are disposed. The output from the OR circuit is used as the abnormal temperature warning signal, while the output from the AND circuit is utilized as a signal for actuating the scram operation of control rod drive mechanisms. Accordingly, it is possible to improve the reliability of the reactor scram system, particularly, improve the reliability under a high temperature atmosphere. (Kamimura, M.)

  7. Safety analysis report for the National Low-Temperature Neutron Irradiation Facility (NLTNIF) at the ORNL Bulk Shielding Reactor (BSR)

    International Nuclear Information System (INIS)

    Coltman, R.R. Jr.; Kerchner, H.R.; Klabunde, C.E.; Richardson, S.A.

    1986-06-01

    This report provides information concerning: the experiment facility; experiment assembly; instrumentation and controls; materials; radioactivity; shielding; thermodynamics; estimated or measured reactivity effects; procedures; hazards; and quality assurance

  8. Lining facility for FBR type reactor

    International Nuclear Information System (INIS)

    Shimano, Kunio.

    1991-01-01

    In a lining facility for protecting structural material concretes for concrete buildings in an FBR type power plant, sodium-resistant and heat-resistant first and second coating layers are lined at the surface of concretes, and steam releasing materials are disposed between the first and the second coating layers for releasing water contents evaporated from the concretes to the outside. With such a constitution, since there is no structures for welding steel plates to each other as in the prior art, the fabrication is made easy. Further, since cracks of coating materials can be suppressed, reactor safety is improved. (T.M.)

  9. New perspectives on reactor safety

    International Nuclear Information System (INIS)

    Avery, R.

    1986-01-01

    Over the past few years a number of changes and new perspectives have come about in our approach to reactor safety. These changes have occurred over a period of time extending from as long ago as 1975, when WASH-1400 came out representing the first major application of probabilistic risk analysis (PRA) to US reactor plants. The period of change has extended from that time to the present, and includes new areas of focus such as safety goals, source term studies, and severe accident policy statement and approaches, including the IDCOR Program. It has also included a greatly increased interest in inherent safety. These areas are discussed in this paper

  10. IRSN research programs concerning reactor safety

    International Nuclear Information System (INIS)

    Bardelay, J.

    2005-01-01

    This paper is made up of 3 parts. The first part briefly presents the missions of IRSN (French research institute on nuclear safety), the second part reviews the research works currently led by IRSN in the following fields : -) the assessment of safety computer codes, -) thermohydraulics, -) reactor ageing, -) reactivity accidents, -) loss of coolant, -) reactor pool dewatering, -) core meltdown, -) vapor explosion, and -) fission product release. In the third part, IRSN is shown to give a major importance to experimental programs led on research or test reactors for collecting valid data because of the complexity of the physical processes that are involved. IRSN plans to develop a research program concerning the safety of high or very high temperature reactors. (A.C.)

  11. Reactor Sharing at Rensselaer Critical Facility

    International Nuclear Information System (INIS)

    D. Steiner, D. Harris, T. Trumbull

    2006-01-01

    This final report summarizes the reactor sharing activities at the Rensselaer Critical Facility. An example of a typical tour is also included. Reactor sharing at the RCF brings outside groups into the facility for a tour, an explanation of reactor matters, and a reactor measurement. It has involved groups ranging from high school classes to advanced college groups and in size from a few to about 50 visitors. The RCF differs from other university reactors in that its fuel is like that of large power reactors, and its research and curriculum are dedicated to power reactor matters

  12. Design requirements for new nuclear reactor facilities in Canada

    International Nuclear Information System (INIS)

    Shim, S.; Ohn, M.; Harwood, C.

    2012-01-01

    The Canadian Nuclear Safety Commission (CNSC) has been establishing the regulatory framework for the efficient and effective licensing of new nuclear reactor facilities. This regulatory framework includes the documentation of the requirements for the design and safety analysis of new nuclear reactor facilities, regardless of size. For this purpose, the CNSC has published the design and safety analysis requirements in the following two sets of regulatory documents: 1. RD-337, Design of New Nuclear Power Plants and RD-310, Safety Analysis for Nuclear Power Plants; and 2. RD-367, Design of Small Reactor Facilities and RD-308, Deterministic Safety Analysis for Small Reactor Facilities. These regulatory documents have been modernized to document past practices and experience and to be consistent with national and international standards. These regulatory documents provide the requirements for the design and safety analysis at a high level presented in a hierarchical structure. These documents were developed in a technology neutral approach so that they can be applicable for a wide variety of water cooled reactor facilities. This paper highlights two particular aspects of these regulatory documents: The use of a graded approach to make the documents applicable for a wide variety of nuclear reactor facilities including nuclear power plants (NPPs) and small reactor facilities; and, Design requirements that are new and different from past Canadian practices. Finally, this paper presents some of the proposed changes in RD-337 to implement specific details of the recommendations of the CNSC Fukushima Task Force Report. Major changes were not needed as the 2008 version of RD-337 already contained requirements to address most of the lessons learned from the Fukushima event of March 2011. (author)

  13. FENIX [Fusion ENgineering International eXperimental]: A test facility for ITER [International Thermonuclear Experimental Reactor] and other new superconducting magnets

    International Nuclear Information System (INIS)

    Slack, D.S.; Patrick, R.E.; Miller, J.R.

    1990-01-01

    The Fusion ENgineering International eXperimental (FENIX) Test Facility which is nearing completion at Lawrence Livermore National Laboratory, is a 76-t set of superconducting magnets housed in a 4-m-diameter cryostat. It represents a significant step toward meeting the testing needs for the development of superconductors appropriate for large-scale magnet applications such as the International Thermonuclear Experimental Reactor (ITER). The magnet set is configured to allow radial access to the 0.4-m-diameter high-field region where maximum fields up to 14 T will be provided. The facility is fitted with a thermally isolated test well with a port to the high-field region that allows insertion and removal of test conductors without disturbing the cryogenic environment of the magnets. It is expected that the facility will be made available to magnet developers internationally, and this paper discusses its general design features, its construction, and its capabilities

  14. Engineering test facility design center

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    The vehicle by which the fusion program would move into the engineering testing phase of fusion power development is designated the Engineering Test Facility (ETF). The ETF would provide a test bed for reactor components in the fusion environment. In order to initiate preliminary planning for the ETF decision, the Office of Fusion Energy established the ETF Design Center activity to prepare the design of the ETF. This section describes the status of this design

  15. Advanced Instrumentation for Transient Reactor Testing

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, Michael L.; Anderson, Mark; Imel, George; Blue, Tom; Roberts, Jeremy; Davis, Kurt

    2018-01-31

    Transient testing involves placing fuel or material into the core of specialized materials test reactors that are capable of simulating a range of design basis accidents, including reactivity insertion accidents, that require the reactor produce short bursts of intense highpower neutron flux and gamma radiation. Testing fuel behavior in a prototypic neutron environment under high-power, accident-simulation conditions is a key step in licensing nuclear fuels for use in existing and future nuclear power plants. Transient testing of nuclear fuels is needed to develop and prove the safety basis for advanced reactors and fuels. In addition, modern fuel development and design increasingly relies on modeling and simulation efforts that must be informed and validated using specially designed material performance separate effects studies. These studies will require experimental facilities that are able to support variable scale, highly instrumented tests providing data that have appropriate spatial and temporal resolution. Finally, there are efforts now underway to develop advanced light water reactor (LWR) fuels with enhanced performance and accident tolerance. These advanced reactor designs will also require new fuel types. These new fuels need to be tested in a controlled environment in order to learn how they respond to accident conditions. For these applications, transient reactor testing is needed to help design fuels with improved performance. In order to maximize the value of transient testing, there is a need for in-situ transient realtime imaging technology (e.g., the neutron detection and imaging system like the hodoscope) to see fuel motion during rapid transient excursions with a higher degree of spatial and temporal resolution and accuracy. There also exists a need for new small, compact local sensors and instrumentation that are capable of collecting data during transients (e.g., local displacements, temperatures, thermal conductivity, neutron flux, etc.).

  16. Operational safety and reactor life improvements of Kyoto University Reactor

    International Nuclear Information System (INIS)

    Utsuro, M.; Fujita, Y.; Nishihara, H.

    1990-01-01

    Recent important experience in improving the operational safety and life of a reactor are described. The Kyoto University Reactor (KUR) is a 25-year-old 5 MW light water reactor provided with two thermal columns of graphite and heavy water as well as other kinds of experimental facilities. In the graphite thermal column, noticeable amounts of neutron irradiation effects had accumulated in the graphite blocks near the core. Before the possible release of the stored energy, all the graphite blocks in the column were successfully replaced with new blocks using the opportunity provided by the installation of a liquid deuterium cold neutron source in the column. At the same time, special seal mechanisms were provided for essential improvements to the problem of radioactive argon production in the column. In the heavy-water thermal column we have accomplished the successful repair of a slow leak of heavy water through a thin instrumentation tube failure. The repair work included the removal and reconstructions of the lead and graphite shielding layers and welding of the instrumentation tube under radiation fields. Several mechanical components in the reactor cooling system were also exchanged for new components with improved designs and materials. On-line data logging of almost all instrumentation signals is continuously performed with a high speed data analysis system to diagnose operational conditions of the reactor. Furthermore, through detailed investigations on critical components, operational safety during further extended reactor life will be supported by well scheduled maintenance programs

  17. Textiles Performance Testing Facilities

    Data.gov (United States)

    Federal Laboratory Consortium — The Textiles Performance Testing Facilities has the capabilities to perform all physical wet and dry performance testing, and visual and instrumental color analysis...

  18. GPS Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Global Positioning System (GPS) Test Facility Instrumentation Suite (GPSIS) provides great flexibility in testing receivers by providing operational control of...

  19. Safety of magnetic fusion facilities: Guidance

    International Nuclear Information System (INIS)

    1996-05-01

    This document provides guidance for the implementation of the requirements identified in DOE-STD-6002-96, Safety of Magnetic Fusion Facilities: Requirements. This guidance is intended for the managers, designers, operators, and other personnel with safety responsibilities for facilities designated as magnetic fusion facilities. While the requirements in DOE-STD-6002-96 are generally applicable to a wide range of fusion facilities, this Standard, DOE-STD-6003-96, is concerned mainly with the implementation of those requirements in large facilities such as the International Thermonuclear Experimental Reactor (ITER). Using a risk-based prioritization, the concepts presented here may also be applied to other magnetic fusion facilities. This Standard is oriented toward regulation in the Department of Energy (DOE) environment as opposed to regulation by other regulatory agencies. As the need for guidance involving other types of fusion facilities or other regulatory environments emerges, additional guidance volumes should be prepared. The concepts, processes, and recommendations set forth here are for guidance only. They will contribute to safety at magnetic fusion facilities

  20. Safety classification of systems, structures, and components for pool-type research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Ryong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2016-08-15

    Structures, systems, and components (SSCs) important to safety of nuclear facilities shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions. Although SSC classification guidelines for nuclear power plants have been well established and applied, those for research reactors have been only recently established by the International Atomic Energy Agency (IAEA). Korea has operated a pool-type research reactor (the High Flux Advanced Neutron Application Reactor) and has recently exported another pool-type reactor (Jordan Research and Training Reactor), which is being built in Jordan. Korea also has a plan to build one more pool-type reactor, the Kijang Research Reactor, in Kijang, Busan. The safety classification of SSCs for pool-type research reactors is proposed in this paper based on the IAEA methodology. The proposal recommends that the SSCs of pool-type research reactors be categorized and classified on basis of their safety functions and safety significance. Because the SSCs in pool-type research reactors are not the pressure-retaining components, codes and standards for design of the SSCs following the safety classification can be selected in a graded approach.

  1. JHR Project: a future Material Testing Reactor working as an International user Facility: The key-role of instrumentation in support to the development of modern experimental capacity

    Energy Technology Data Exchange (ETDEWEB)

    Bignan, G. [CEA, DEN, DER, JHR user Facility Interface Manager' , Cadarache, F-13108 St-Paul-Lez-Durance (France); Gonnier, C. [CEA, DEN, DER, SRJH Jules Horowitz Reactor Service, Cadarache, F-13108 St-Paul-Lez-Durance (France); Lyoussi, A.; Villard, J.F.; Destouches, C. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France); Chauvin, J.P. [CEA,DEN, DER, SPEX, Experimental Physics Service, Cadarache, F-13108 St-Paul-Lez-Durance (France); Maugard, B. [CEA, DEN, DER, Reactor Department Studies, Cadarache, F-13108 St-Paul-Lez-Durance (France)

    2015-07-01

    Research and development on fuel and material behaviour under irradiation is a key issue for sustainable nuclear energy in order to meet specific needs by keeping the best level of safety. These needs mainly deal with a constant improvement of performances and safety in order to optimize the fuel cycle and hence to reach nuclear energy sustainable objectives. A sustainable nuclear energy requires a high level of performances in order to meet specific needs such as: - Pursuing improvement of the performances and safety of present and coming water cooled reactor technologies. This will require a continuous R and D support following a long-term trend driven by the plant life management, safety demonstration, flexibility and economics improvement. Experimental irradiations of structure materials are necessary to anticipate these material behaviours and will contribute to their optimisation. - Upgrading continuously nuclear fuel technology in present and future nuclear power plants to achieve better performances and to optimise the fuel cycle keeping the best level of safety. Fuel evolution for generation II, III and III+ is a key stake requiring developments, qualification tests and safety experiments to ensure the competitiveness and safety: experimental tests exploring the full range of fuel behaviour determine fuel stability limits and safety margins, as a major input for the fuel reliability analysis. To perform such accurate and innovative progress and developments, specific and ad hoc instrumentation, irradiation devices, measurement methods are necessary to be set up inside or beside the material testing reactor (MTR) core. These experiments require beforehand in situ and on line sophisticated measurements to accurately determine different key parameters such as thermal and fast neutron fluxes and nuclear heating in order to precisely monitor and control the conducted assays. The new Material Testing Reactor JHR (Jules Horowitz Reactor) currently under

  2. Safety evaluation report related to operation of Fast Flux Test Facility. Supplement No. 1

    International Nuclear Information System (INIS)

    1979-05-01

    This supplement provides (1) the staff's evaluation of additional information received since issuance of the Safety Evaluation Report regarding previously identified uncompleted review items, (2) a discussion of comments made by the ACRS in its report of November 8, 1978, and (3) the staff's evaluation of additional or revised information related to new or old issues that have arisen since the issuance of the Safety Evaluation Report

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

    International Nuclear Information System (INIS)

    Konzek, G.J.

    1983-07-01

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

  4. Thermal reactor safety

    International Nuclear Information System (INIS)

    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

  5. Thermal reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    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.

  6. N Reactor updated safety analysis report, NUSAR

    International Nuclear Information System (INIS)

    1978-01-01

    An update of the N Reactor safety analysis is presented to reconfirm that the continued operation does not pose undue risk to DOE personnel and property, the public, or the environment. A reanalysis of LOCA and reactivity transients utilizing current codes and methods is made. The principal aspects of the overall submission, a general description, and site characteristics including geography and demography, nearby industrial, transportation and military facilities, meteorology, hydraulic engineering, and geology and seismology are described

  7. CLEAR test facility

    CERN Multimedia

    Ordan, Julien Marius

    2017-01-01

    A new user facility for accelerator R&D, the CERN Linear Electron Accelerator for Research (CLEAR), started operation in August 2017. CLEAR evolved from the former CLIC Test Facility 3 (CTF3) used by the Compact Linear Collider (CLIC). The new facility is able to host and test a broad range of ideas in the accelerator field.

  8. Application of ultrasonic testing technique to detect gas accumulation in important pipings for pressurized water reactors safety

    Energy Technology Data Exchange (ETDEWEB)

    Fushimi, Yasuyuki [Inst. of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2002-09-01

    Since 1988, the USNRC has pointed out that gas-binding events might occur at high head safety injection (HHSI) pumps of pressurized water reactors (PWRs). In Japanese PWR plants, corrective actions were taken in response to gas-binding events that occurred on HHSI pumps in the USA, so no gas accumulation event has been reported so far. However, when venting frequency is prolonged with operating cycle extension, the probability of gas accumulation in pipings may increase as in the USA. The purpose of this study was to establish a technique to identify gas accumulation and to measure the gas volume accurately. Taking dominant causes of the gas-binding events in the USA into consideration, we pointed out the following sections in the Japanese PWRs where gas srtipping and/or gas accumulation might occur: residual heat removal system pipings and charging/safety injection pump minimum flow line. Then an ultrasonic testing technique, adopted to identify gas accumulation in the USA, was applied to those sections of the typical Japanese PWR. Consequently, no gas accumulation was found in those pipings. (author)

  9. Overview of fusion reactor safety

    International Nuclear Information System (INIS)

    Cohen, S.; Crocker, J.G.

    1981-01-01

    Use of deuterium-tritium burning fusion reactors requires examination of several major safety and environmental issues: (1) tritium inventory control, (2) neutron activation of structural materials, fluid streams and reactor hall environment, (3) release of radioactivity from energy sources including lithium spill reactions, superconducting magnet stored energy release, and plasma disruptions, (4) high magnetic and electromagnetic fields associated with fusion reactor superconducting magnets and radio frequency heating devices, and (5) handling and disposal of radioactive waste. Early recognition of potential safety problems with fusion reactors provides the opportunity for improvement in design and materials to eliminate or greatly reduce these problems. With an early start in this endeavor, fusion should be among the lower risk technologies for generation of commercial electrical power

  10. Ouellette Thermal Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Thermal Test Facility is a joint Army/Navy state-of-the-art facility (8,100 ft2) that was designed to:Evaluate and characterize the effect of flame and thermal...

  11. The Self-Calibration Test of flowmeter installed in STELLA(Sodium Integral Effect Test Loop for Safety Simulation and Assessment) facility

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Minhwan; Jeong, Ji-Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The objective of this study is to describe the procedure of the self-calibration test for the flowmeters and to analyze the result of the test. In this work, the test procedure of the self-calibration of two flowmeters (FT-101, FT-102) installed in STELLA facility was described and the test result was analyzed. In regard to the long-term SFR development plan, a large-scale sodium thermal-hydraulic test project is being progressed by KAERI. This project is called STELLA (Sodium Integral Effect Test Loop for Safety Simulation and Assessment), and it is proceeding by adopting the QA (Quality Assurance) program. Due to the specificity of an experiment using sodium(Na) categorized as Class 3(pyrophoric material and water-prohibiting substance) by the Safety Control of Dangerous Substances Act, it is necessary to apply QA in consideration of the sodium experiment environment in certain parts. The one of them is about calibration of measuring instrument such as a flowmeter, thermocouple and pressure gauge. It is described in the QAP (Quality Assurance Procedures) of KAERI that calibration work should be conducted in accordance with self-calibration procedures in a special case where conventional calibration is not practicable. The calibration of two flowmeters (FT-101, FT-102) installed in STELLA facility is the typical example. As a result of test, it was confirmed that the flowmeters meet the pass criterion. Therefore, it was concluded that the flowmeters maintain instrument capacity a year ago.

  12. Spontaneous stabilization of HTGRs without reactor scram and core cooling—Safety demonstration tests using the HTTR: Loss of reactivity control and core cooling

    Energy Technology Data Exchange (ETDEWEB)

    Takamatsu, Kuniyoshi, E-mail: takamatsu.kuniyoshi@jaea.go.jp; Yan, Xing L.; Nakagawa, Shigeaki; Sakaba, Nariaki; Kunitomi, Kazuhiko

    2014-05-01

    It is well known that a High-Temperature Gas-cooled Reactor (HTGR) has superior safety characteristics; for example, an HTGR has a self-control system that uses only physical phenomena against various accidents. Moreover, the large heat capacity and low power density of the core result in very slow temperature transients. Therefore, an HTGR serves inherently safety features against loss of core cooling accidents such as the Tokyo Electric Power Co., Inc. (TEPCO)’s Fukushima Daiichi Nuclear Power Station (NPS) disaster. Herein we would like to demonstrate the inherent safety features using the High-Temperature Engineering Test Reactor (HTTR). The HTTR is the first HTGR in Japan with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950 °C; it was built at the Oarai Research and Development Center of Japan Atomic Energy Agency (JAEA). In this study, an all-gas-circulator trip test was analyzed as a loss of forced cooling (LOFC) test with an initial reactor power of 9 MW to demonstrate LOFC accidents. The analytical results indicate that reactor power decreases from 9 MW to 0 MW owing to the negative reactivity feedback effect of the core, even if the reactor shutdown system is not activated. The total reactivity decreases for 2–3 h and then gradually increases in proportion to xenon reactivity; therefore, the HTTR achieves recritical after an elapsed time of 6–7 h, which is different from the elapsed time at reactor power peak occurrence. After the reactor power peak occurs, the total reactivity oscillates several times because of the negative reactivity feedback effect and gradually decreases to zero. Moreover, the new conclusions are as follows: the greater the amount of residual heat removed from the reactor core, the larger the stable reactor power after recriticality owing to the heat balance of the reactor system. The minimum reactor power and the reactor power peak occurrence are affected by the neutron source. The greater the

  13. Integral fast reactor safety features

    International Nuclear Information System (INIS)

    Cahalan, J.E.; Kramer, J.M.; Marchaterre, J.F.; Mueller, C.J.; Pedersen, D.R.; Sevy, R.H.; Wade, D.C.; Wei, T.Y.C.

    1988-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. In addition to liquid metal cooling, the principal design features that distinguish the IFR are: (1) a pool-type primary system, (2) an advanced ternary alloy metallic fuel, and (3) an integral fuel cycle with on-site fuel reprocessing and fabrication. This paper focuses on the technical aspects of the improved safety margins available in the IFR concept. This increased level of safety is made possible by (1) the liquid metal (sodium) coolant and pool-type primary system layout, which together facilitate passive decay heat removal, and (2) a sodium-bonded metallic fuel pin design with thermal and neutronic properties that provide passive core responses which control and mitigate the consequences of reactor accidents

  14. Integral fast reactor safety features

    International Nuclear Information System (INIS)

    Cahalan, J.E.; Kramer, J.M.; Marchaterre, J.F.; Mueller, C.J.; Pedersen, D.R.; Sevy, R.H.; Wade, D.C.; Wei, T.Y.C.

    1988-01-01

    The integral fast reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. In addition to liquid metal cooling, the principal design features that distinguish the IFR are: a pool-type primary system, and advanced ternary alloy metallic fuel, and an integral fuel cycle with on-site fuel reprocessing and fabrication. This paper focuses on the technical aspects of the improved safety margins available in the IFR concept. This increased level of safety is made possible by the liquid metal (sodium) coolant and pool-type primary system layout, which together facilitate passive decay heat removal, and a sodium-bonded metallic fuel pin design with thermal and neutronic properties that provide passive core responses which control and mitigate the consequences of reactor accidents

  15. Refurbishment and safety up-gradation of Cirus Reactor

    International Nuclear Information System (INIS)

    Rao, D.V.H.

    2004-01-01

    Cirus, a 40 MWth, vertical tank type research reactor, having a wide range of research facilities, was commissioned in 1960. This research facility has been operated and utilized extensively for nearly four decades. With a view to assess the residual life of the reactor, detailed ageing studies were carried out. Based on this, refurbishment work for life extension was undertaken. During this work, additional safety features were incorporated to improve the overall safety of the reactor. This lecture details the methodologies used for ageing studies and refurbishment activities for life extension with enhanced safety. (author)

  16. Cadmium safety rod thermal tests

    International Nuclear Information System (INIS)

    Thomas, J.K.; Iyer, N.C.; Peacock, H.B.

    1992-01-01

    Thermal testing of cadmium safety rods was conducted as part of a program to define the response of Savannah River Site (SRS) production reactor core components to a hypothetical LOCA leading to a drained reactor tank. The safety rods are present in the reactor core only during shutdown and are not used as a control mechanism during operation; thus, their response to the conditions predicted for the LOCA is only of interest to the extent that it could impact the progression of the accident. This document provides a description of this testing

  17. Thermal Reactor Safety

    International Nuclear Information System (INIS)

    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

  18. Thermal Reactor Safety

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. Department of Energy's High Flux Isotope Reactor (HFIR), October 20--24, 1980: A special report prepared for the Nuclear Facilities Personnel Qualification and Training Committee: An independent on-site safety review

    International Nuclear Information System (INIS)

    1981-02-01

    The intent of this on-site safety review was to make a broad management assessment of HFIR operations, rather than conduct a detailed in-depth audit. The result of the review should only be considered as having identified trends or indications. The Team's observations and recommendations are based upon licensed reactor facility practices used to meet industry standards. For the most part, these standards form the basis for many of the comments in this report. The Team believes that a uniform minimum standard of performance should be achieved in the operation of DOE reactors. In order to assure that this is accomplished, clear standards are necessary. Consistent with the provisions of past AEC and ERDA policy, the Team has used the standards of the commercial nuclear power industry. It is recognized that this approach is conservative in that the HFIR reactor has a significantly greater degree of inherent safety (low temperature, low pressure, low power) than a licensed reactor

  20. The high flux reactor Petten, A multi-purpose research and test facility for the future of nuclear energy

    International Nuclear Information System (INIS)

    Bergmans, H.; Duijves, K.; Conrad, R.; Markgraf, J.F.W.; May, R.; Moss, R.L.; Sordon, G.; Tartaglia, G.P.

    1996-01-01

    The High Flux Reactor (HFR) at Petten, is owned by the European Commission (EC) and managed by the Institute for Advanced Materials (IAM) of the Joint Research Centre (JRC) of the EC. Its operation has been entrusted since 1962 to the Netherlands Energy Research Foundation (ECN). The HFR is one of the most powerful multi-purpose research and test reactors in the world. Together with the ECN hot cells at Petten, it has provided since three decades an integral and full complement of irradiation and examination services as required by current and future research and development for nuclear energy, industry and research organizations. Since 1963, the HFR has recognized record of consistent, reliable and high availability of more than 250 days of operation per year. The HFR has 20 in-core and 12 poolside irradiation positions, plus 12 beam tubes. With a variety of dedicated irradiation devices, and with its long-standing experience in executing small and large irradiation projects, the HFR is particularly suited for fuel, materials and components testing for all reactor lines, including thermonuclear fusion reactors. In addition, processing with neutrons and gamma rays, neutron-based research and inspection services are employed by industry and research, such as activation analysis, boron neutron capture therapy, neutron radiography and neutron diffraction. Moreover, in recent years, HFRs' mission has been broadened within the area of radioisotopes production, where, within a few years, the HFR has attained the European leadership in production volume

  1. Development of safety analysis technology for integral reactor

    International Nuclear Information System (INIS)

    Kim, Hee Cheol; Kim, K. K.; Kim, S. H.

    2002-04-01

    The state-of-the-arts for the integral reactor was performed to investigate the safety features. The safety and performance of SMART were assessed using the technologies developed during the study. For this purpose, the computer code system and the analysis methodology were developed and the safety and performance analyses on SMART basic design were carried out for the design basis event and accident. The experimental facilities were designed for the core flow distribution test and the self-pressurizing pressurizer performance test. The tests on the 2-phase critical flow with non-condensable gas were completed and the results were used to assess the critical flow model. Probabilistic Safety Assessment(PSA) was carried out to evaluate the safety level and to optimize the design by identifying and remedying any weakness in the design. A joint study with KINS was carried out to promote licensing environment. The generic safety issues of integral reactors were identified and the solutions were formulated. The economic evaluation of the SMART desalination plant and the activities related to the process control were carried out in the scope of the study

  2. Research reactor management. Safety improvement activities in HANARO

    International Nuclear Information System (INIS)

    Wu, Jong-Sup; Jung, Hoan-Sung; Hong, Sung Taek; Ahn, Guk-Hoon

    2012-01-01

    Safety activities in HANARO have been continuously conducted to enhance its safe operation. Great effort has been placed on a normalization and improvement of the safety attitude of the regular staff and other employees working at the reactor and other experimental facilities. This paper introduces the activities on safety improvement that were performed over the last few years. (author)

  3. Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility

    International Nuclear Information System (INIS)

    1992-07-01

    Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility

  4. Licensing procedures and safety criteria for research reactors in France

    International Nuclear Information System (INIS)

    Berry, J.L.; Lerouge, B.

    1980-11-01

    This paper summarizes the recent evolution of the French research reactor capacity, describes the licensing process, the main safety criteria which are taken into consideration, and associated safety research. Some of the existing facilities underwent important modifications to comply with more severe safety criteria, increase the experimental capabilities or qualify new low-enrichment fuels for research reactors. At the end, a few considerations are given to the consequences of the Osiris core conversion

  5. Research for enhancing reactor safety

    International Nuclear Information System (INIS)

    1989-05-01

    Recent research for enhanced reactor safety covers extensive and numerous experiments and computed modelling activities designed to verify and to improve existing design requirements. The lectures presented at the meeting report GRS research results and the current status of reactor safety research in France. The GRS experts present results concerning expert systems and their perspectives in safety engineering, large-scale experiments and their significance in the development and verification of computer codes for thermohydraulic modelling of safety-related incidents, the advanced system code ATHLET for analysis of thermohydraulic processes of incidents, the analysis simulator which is a tool for fast evaluation of accident management measures, and investigations into event sequences and the required preventive emergency measures within the German Risk Study. (DG) [de

  6. Nuclear Reactor RA Safety Report, Vol. 14, Safety protection measures

    International Nuclear Information System (INIS)

    1986-11-01

    Nuclear reactor accidents can be caused by three type of errors: failure of reactor components including (1) control and measuring instrumentation, (2) errors in operation procedure, (3) natural disasters. Safety during reactor operation are secured during its design and construction and later during operation. Both construction and administrative procedures are applied to attain safe operation. Technical safety features include fission product barriers, fuel elements cladding, primary reactor components (reactor vessel, primary cooling pipes, heat exchanger in the pump), reactor building. Safety system is the system for safe reactor shutdown and auxiliary safety system. RA reactor operating regulations and instructions are administrative acts applied to avoid possible human error caused accidents [sr

  7. Mark 1 Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Mark I Test Facility is a state-of-the-art space environment simulation test chamber for full-scale space systems testing. A $1.5M dollar upgrade in fiscal year...

  8. Structural Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Provides a wide variety of testing equipment, fixtures and facilities to perform both unique aviation component testing as well as common types of materials testing...

  9. Environmental concerns in regarding a materials test reactor fuel fabrication facility at the Nuclear and Energy Research Institute - IPEN

    International Nuclear Information System (INIS)

    Santos, Glaucia R.T.; Durazzo, Michelangelo; Carvalho, Elita F.U.; Riella, Humberto G.

    2008-01-01

    The aim of the industrial activities success, front to a more and more informed and demanding society and to a more and more competitive market demands an environmental administration policy which doesn't limit itself to assist the legislation but anticipate and prevent, in a responsible way, possible damages to the environment. One of the main programs of the Institute of Energetic and Nuclear Research of the national Commission of Nuclear Energy located in Brazil, through the Center of Nuclear Fuel -CCN- is to manufacture MTR-type fuel elements using low-enrichment uranium (20 wt % 235 U), to supply its IEA-R1 research reactor. Integrated in this program, this work aims at well developing and assuring a methodology to implant an environment, health and safety policy, foreseeing its management with the use of detailed data reports and through the adoption of new tools for improving the management, in order to fulfil the applicable legislation and accomplish all the environmental, operational and works aspects. The applied methodology for the effluents management comprises different aspects, including the specific environmental legislation of a country, main available effluents treatment techniques, process flow analyses from raw materials and intakes to products, generated effluents, residuals and emissions. Data collections were accomplished for points gathering and tests characterization, classification and compatibility of the generated effluents and their eventual environmental impacts.This study aims to implant the Sustainability Concept in order to guarantee access to financial resources, allowing cost reduction, maximizing long-term profits, preventing and reducing environmental accident risks and stimulating both the attraction and the keeping of a motivated manpower. Work on this project has already started and, even though many technical actions have not still ended, the results have being extremely valuable. These results can already give to CCN

  10. Environmental concerns regarding a materials test reactor fuel fabrication facility at the Nuclear and Energy Research Institute - IPEN

    International Nuclear Information System (INIS)

    Santos, G. R. T.; Durazzo, M.; Carvalho, E. F. U.; Riella, H. G.

    2008-01-01

    The aim of the industrial activities success, front to a more and more informed and demanding society and to a more and more competitive market demands an environmental administration policy which doesn't limit itself to assist the legislation but anticipate and prevent, in a responsible way, possible damages to the environment. One of the maim programs of the Institute of Energetic and Nuclear Research of the national Commission of Nuclear Energy located in Brazil, through the Center of Nuclear Fuel - CCN - is to manufacture MTR-type fuel elements using low-enrichment uranium (20 wt% 2 35U), to supply its IEA-RI research reactor. Integrated in this program, this work aims at well developing and assuring a methodology to implant an environment, health and safety policy, foreseeing its management with the use of detailed data reports and through the adoption of new tools for improving the management, in order to fulfil the applicable legislation and accomplish all the environmental, operational and works aspects. The applied methodology for the effluents management comprises different aspects, including the specific environmental legislation of a country, main available effluents treatment techniques, process flow analyses from raw materials and intakes to products, generated effluents, residuals and emissions. Data collections were accomplished for points gathering and tests characterization, classification and compatibility of the generated effluents and their eventual environmental impacts. This study aims to implant the Sustainable Concept in order to guarantee access to financial resources, allowing cost reduction, maximizing long-term profits, preventing and reducing environmental accident risks and stimulating both the attraction and the keeping of a motivated manpower. Work on this project has already started and, even though many technical actions have not still ended, the results have being extremely valuable. These results can already give to CCN

  11. Shear compression testing of glass-fibre steel specimens after 4K reactor irradiation: Present status and facility upgrade

    International Nuclear Information System (INIS)

    Gerstenberg, H.; Kraehling, E.; Katheder, H.

    1997-01-01

    The shear strengths of various fibre reinforced resins being promising candidate insulators for superconducting coils to be used tinder a strong radiation load, e.g. in future fusion reactors were investigated prior and subsequent to reactor in-core irradiation at liquid helium temperature. A large number of sandwich-like (steel-bonded insulation-steel) specimens representing a widespread variety of materials and preparation techniques was exposed to irradiation doses of up to 5 x 10 7 Gy in form of fast neutrons and γ-radiation. In a systematic study several experimental parameters including irradiation dose, postirradiation storage temperature and measuring temperature were varied before the determination of the ultimate shear strength. The results obtained from the different tested materials are compared. In addition an upgrade of the in-situ test rig installed at the Munich research reactor is presented, which allows combined shear/compression loading of low temperature irradiated specimens and provides a doubling of the testing rate

  12. Pavement Testing Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Comprehensive Environmental and Structural AnalysesThe ERDC Pavement Testing Facility, located on the ERDC Vicksburg campus, was originally constructed to provide an...

  13. Guide On Safety Tests

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1987-05-15

    This book tells US important things to do safety tests, which includes basic caution for experiment treatment of used materials such as ignition substance inflammables, explosive substance and toxic substance, handling of used equipment like inflammable device, machine, high pressure device, high pressure gas, and high energy device, first aid such as addiction by drug, flame, radiation exposure, and heart massage treatment of waste in laboratory like cautions on general treatment, handling of inorganic waste, organic waste and waste treatment with disposal facilities.

  14. Guide On Safety Tests

    International Nuclear Information System (INIS)

    1987-05-01

    This book tells US important things to do safety tests, which includes basic caution for experiment treatment of used materials such as ignition substance inflammables, explosive substance and toxic substance, handling of used equipment like inflammable device, machine, high pressure device, high pressure gas, and high energy device, first aid such as addiction by drug, flame, radiation exposure, and heart massage treatment of waste in laboratory like cautions on general treatment, handling of inorganic waste, organic waste and waste treatment with disposal facilities.

  15. Nuclear reactor safety in the USA

    International Nuclear Information System (INIS)

    Vigil, J.C.

    1983-01-01

    Nuclear reactor safety in the USA has emphasized a defense-in-depth approach to protecting the public from reactor accidents. This approach was severely tested by the Three Mile Island accident and was found to be effective in safeguarding the public health and safety. However, the economic impact of the TMI accident was very large. Consequently, more attention is now being given to plant protection as well as public-health protection in reactor-safety studies. Sophisticated computer simulations at Los Alamos are making major contributions in this area. In terms of public risk, nuclear power plants compare favorably with other large-scale alternatives to electricity generation. Unfortunately, there is a large gulf between the real risks of nuclear power and the present public perception of these risks

  16. Operating experience of steam generator test facility

    International Nuclear Information System (INIS)

    Sureshkumar, V.A.; Madhusoodhanan, G.; Noushad, I.B.; Ellappan, T.R.; Nashine, B.K.; Sylvia, J.I.; Rajan, K.K.; Kalyanasundaram, P.; Vaidyanathan, G.

    2006-01-01

    Steam Generator (SG) is the vital component of a Fast Reactor. It houses both water at high pressure and sodium at low pressure separated by a tube wall. Any damage to this barrier initiates sodium water reaction that could badly affect the plant availability. Steam Generator Test Facility (SGTF) has been set up in Indira Gandhi Centre for Atomic Research (IGCAR) to test sodium heated once through steam generator of 19 tubes similar to the PFBR SG dimension and operating conditions. The facility is also planned as a test bed to assess improved designs of the auxiliary equipments used in Fast Breeder Reactors (FBR). The maximum power of the facility is 5.7 MWt. This rating is arrived at based on techno economic consideration. This paper covers the performance of various equipments in the system such as Electro magnetic pumps, Centrifugal sodium pump, in-sodium hydrogen meters, immersion heaters, and instrumentation and control systems. Experience in the system operation, minor modifications, overall safety performance, and highlights of the experiments carried out etc. are also brought out. (author)

  17. Analysis of an Advanced Test Reactor Small-Break Loss-of-Coolant Accident with an Engineered Safety Feature to Automatically Trip the Primary Coolant Pumps

    International Nuclear Information System (INIS)

    Polkinghorne, Steven T.; Davis, Cliff B.; McCracken, Richard T.

    2000-01-01

    A new engineered safety feature that automatically trips the primary coolant pumps following a low-pressure reactor scram was recently installed in the Advanced Test Reactor (ATR). The purpose of this engineered safety feature is to prevent the ATR's surge tank, which contains compressed air, from emptying during a small-break loss-of-coolant accident (SBLOCA). If the surge tank were to empty, the air introduced into the primary coolant loop could potentially cause the performance of the primary and/or emergency coolant pumps to degrade, thereby reducing core thermal margins. Safety analysis performed with the RELAP5 thermal-hydraulic code and the SINDA thermal analyzer shows that adequate thermal margins are maintained during an SBLOCA with the new engineered safety feature installed. The analysis also shows that the surge tank will not empty during an SBLOCA even if one of the primary coolant pumps fails to trip

  18. Reactor containment and reactor safety in the United States

    International Nuclear Information System (INIS)

    Kouts, H.

    1986-01-01

    The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.) [pt

  19. Mirror Fusion Test Facility magnet

    International Nuclear Information System (INIS)

    Henning, C.H.; Hodges, A.J.; Van Sant, J.H.; Hinkle, R.E.; Horvath, J.A.; Hintz, R.E.; Dalder, E.; Baldi, R.; Tatro, R.

    1979-01-01

    The Mirror Fusion Test Facility (MFTF) is the largest of the mirror program experiments for magnetic fusion energy. It seeks to combine and extend the near-classical plasma confinement achieved in 2XIIB with the most advanced neutral-beam and magnet technologies. The product of ion density and confinement time will be improved more than an order of magnitude, while the superconducting magnet weight will be extrapolated from the 15 tons in Baseball II to 375 tons in MFTF. Recent reactor studies show that the MFTF will traverse much of the distance in magnet technology towards the reactor regime. Design specifics of the magnet are given

  20. Nuclear reactor safety system

    International Nuclear Information System (INIS)

    Sato, Takashi.

    1979-01-01

    Purpose: To allow sufficient removal of radioactive substance released in the reactor containment shell upon loss of coolants accidents thus to sufficiently decrease the exposure dose to human body. Constitution: A clean-up system is provided downstream of a heat exchanger and it is branched into a pipeway to be connected to a spray nozzle and further connected by way of a valve to a reactor container. After the end of sudden transient changes upon loss of coolants accidents, the pool water stored in the pressure suppression chamber is purified in the clean-up system and then sprayed in the dry-well by way of a spray nozzle. The sprayed water dissolves to remove water soluble radioactive substances floating in the dry-well and then returns to the pressure suppression chamber. Since radioactive substances in the dry-well can thus removed rapidly and effectively and the pool water can be reused, public hazard can also be decreased. (Horiuchi, T.)

  1. Fire safety requirements for electrical cables towards nuclear reactor safety

    International Nuclear Information System (INIS)

    Raju, M.R.

    2002-01-01

    Full text: Electrical power supply forms a very important part of any nuclear reactor. Power supplies have been categorized in to class I, II, III and IV from reliability point. The safety related equipment are provided with highly reliable power supply to achieve the safety of very high order. Vast network of cables in a nuclear reactor are grouped and segregated to ensure availability of power to at least one group under all anticipated occurrences. Since fire can result in failures leading to unavailability of power caused by common cause, both passive and active fire protection methods are adopted in addition to fire detection system. The paper describes the requirement for passive fire protection to electrical cables viz. fire barrier and fire breaks. The paper gives an account of the tests required to standardize the products. Fire safety implementation for cables in research reactors is described

  2. Present status of decommissioning in the Musashi Reactor Facility (4)

    International Nuclear Information System (INIS)

    Uchiyama, Takafumi; Tanzawa, Tomio; Mitsuhashi, Ishi; Morishima, Kayoko; Matsumoto, Tetsuo

    2012-01-01

    The decommissioning of the Musashi reactor was decided in 2003. Permanent shutdown of the reactor and stopping the operational functions were conducted in 2004. Transportation of the spent fuels was finished in 2006. After 2007, the system and equipment stopping the functions were stored as installed in the reactor facility as radioactive wastes. After separating nonradioactive wastes such as concretes from radioactive wastes with a contamination test, stopping the functions of liquid waste management facility was performed with newly installed drainage facility for radioisotope use in 2010. Solid waste management facility was also dismantled and removed in the same way as liquid waste management facility in 2011. Radioactive wastes packed in containers were moved and stored in the reactor facility. (T. Tanaka)

  3. Decision of the Council of State on the general regulations for the safety of a disposal facility for reactor waste (398/91)

    International Nuclear Information System (INIS)

    1991-01-01

    These Regulations contain provisions for the safe planning, construction and use of final radioactive waste storage facilities. Licensees with a waste management obligation are responsible for ensuring the safety of such facilities. The Regulations entered into force on 1 March 1991. (NEA)

  4. The safety of light water reactors

    International Nuclear Information System (INIS)

    Pershagen, B.

    1986-04-01

    The book describes the principles and practices of reactor safety as applied to the design, regulation and operation of both pressurized water reactors and boiling water reactors. The central part of the book is devoted to methods and results of safety analysis. Some significant events are described, notably the Three Mile Island accident. The book concludes with a chapter on the PIUS principle of inherent reactor safety as applied to the SECURE type of reactor developed in Sweden. (G.B.)

  5. Material accountancy and control practice at a research reactor facility

    International Nuclear Information System (INIS)

    Bouchard, J.; Maurel, J.J.; Tromeur, Y.

    1982-01-01

    This session surveys the regulations, organization, and accountancy practice that compose the French State System of Accountancy and Control. Practical examples are discussed showing how inventories are verified at a critical assembly facility and at a materials testing reactor

  6. Reactor safety research and development in Chalk River Laboratories

    Energy Technology Data Exchange (ETDEWEB)

    Nitheanandan, T. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    Atomic Energy of Canada Limited's Chalk River Laboratories provides three different services to stakeholders and customers. The first service provided by the laboratory is the implementation of Research and Development (R&D) programs to provide the underlying technological basis of safe nuclear power reactor designs. A significant portion of the Canadian R&D capability in reactor safety resides at Atomic Energy of Canada Limited's Chalk River Laboratories, and this capability was instrumental in providing the science and technology required to aid in the safety design of CANDU power reactors. The second role of the laboratory has been in supporting nuclear facility licensees to ensure the continued safe operation of nuclear facilities, and to develop safety cases to justify continued operation. The licensing of plant life extension is a key industry objective, requiring extensive research on degradation mechanisms, such that safety cases are based on the original safety design data and valid and realistic assumptions regarding the effect of ageing and management of plant life. Recently, Chalk River Laboratories has been engaged in a third role in research to provide the technical basis and improved understanding for decision making by regulatory bodies. The state-of-the-art test facilities in Chalk River Laboratories have been contributing to the R&D needs of all three roles, not only in Canada but also in the international community, thorough Canada's participation in cooperative programs lead by International Atomic Energy Agency and the OECD's Nuclear Energy Agency. (author)

  7. PHOEBUS/UHTREX: a preliminary study of a low-cost facility for transient tests of LMFBR fuel

    International Nuclear Information System (INIS)

    Kirk, W.L.

    1976-08-01

    The results of a brief preliminary design study of a facility for transient nuclear tests of fast breeder reactor fuel are described. The study is based on the use of a reactor building originally built for the UHTREX reactor, and the use of some reactor hardware and reactor design and fabrication technology remaining from the Phoebus-2 reactor of the Rover nulcear rocket propulsion program. The facility is therefore currently identified as the PHOEBUS/UHTREX facility. This facility is believed capable of providing early information regarding fast reactor core accident energetics issues which will be very valuable to the overall LMFBR safety program. Facility performance in conjunction with a reference 127-fuel pin experiment is described. Low cost and early availability of the facility were emphasized in the selection of design features and parameters

  8. PHOEBUS/UHTREX: a preliminary study of a low-cost facility for transient tests of LMFBR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kirk, W.L. (comp.)

    1976-08-01

    The results of a brief preliminary design study of a facility for transient nuclear tests of fast breeder reactor fuel are described. The study is based on the use of a reactor building originally built for the UHTREX reactor, and the use of some reactor hardware and reactor design and fabrication technology remaining from the Phoebus-2 reactor of the Rover nulcear rocket propulsion program. The facility is therefore currently identified as the PHOEBUS/UHTREX facility. This facility is believed capable of providing early information regarding fast reactor core accident energetics issues which will be very valuable to the overall LMFBR safety program. Facility performance in conjunction with a reference 127-fuel pin experiment is described. Low cost and early availability of the facility were emphasized in the selection of design features and parameters.

  9. DRY TRANSFER FACILITY CRITICALITY SAFETY CALCULATIONS

    International Nuclear Information System (INIS)

    C.E. Sanders

    2005-01-01

    This design calculation updates the previous criticality evaluation for the fuel handling, transfer, and staging operations to be performed in the Dry Transfer Facility (DTF) including the remediation area. The purpose of the calculation is to demonstrate that operations performed in the DTF and RF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Dry Transfer Facility Description Document'' (BSC 2005 [DIRS 173737], p. 3-8). A description of the changes is as follows: (1) Update the supporting calculations for the various Category 1 and 2 event sequences as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2005 [DIRS 171429], Section 7). (2) Update the criticality safety calculations for the DTF staging racks and the remediation pool to reflect the current design. This design calculation focuses on commercial spent nuclear fuel (SNF) assemblies, i.e., pressurized water reactor (PWR) and boiling water reactor (BWR) SNF. U.S. Department of Energy (DOE) Environmental Management (EM) owned SNF is evaluated in depth in the ''Canister Handling Facility Criticality Safety Calculations'' (BSC 2005 [DIRS 173284]) and is also applicable to DTF operations. Further, the design and safety analyses of the naval SNF canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. Also, note that the results for the Monitored Geologic Repository (MGR) Site specific Cask (MSC) calculations are limited to the

  10. Ballistic Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Ballistic Test Facility is comprised of two outdoor and one indoor test ranges, which are all instrumented for data acquisition and analysis. Full-size aircraft...

  11. Corrosion Testing Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Corrosion Testing Facility is part of the Army Corrosion Office (ACO). It is a fully functional atmospheric exposure site, called the Corrosion Instrumented Test...

  12. Environmental Test Facility (ETF)

    Data.gov (United States)

    Federal Laboratory Consortium — The Environmental Test Facility (ETF) provides non-isolated shock testing for stand-alone equipment and full size cabinets under MIL-S-901D specifications. The ETF...

  13. Full scale BWR containment LOCA response test at the INKA test facility

    International Nuclear Information System (INIS)

    Wagner, Thomas; Leyer, Stephan

    2015-01-01

    KERENA is an innovative boiling water reactor concept with passive safety systems (Generation III+) of AREVA. The reactor is an evolutionary design of operating BWRs (Generation II). In order to verify the functionality and performance of the KERENA safety concept required for the transient and accident management, the test facility “Integral Teststand Karlstein” (INKA) was built at Karlstein (Germany). It is a mock-up of the KERENA boiling water reactor containment, with integrated pressure suppression system. The complete chain of passive safety components is available. The passive components and the levels are represented in full scale. The volume scaling of the containment compartments is approximately 1:24. The reactor pressure vessel (RPV) is simulated via the steam accumulator of the Karlstein Large Valve Test Facility. This vessel provides an energy storage capacity of approximately 1/6 of the KERENA RPV and is supplied by a Benson boiler with a thermal power of 22 MW. With respect to the available power supply, the containment- and system-sizing of the facility is by far the largest one of its kind worldwide. From 2009 to 2012, several single component tests were conducted (Emergency Condenser, Containment Cooling Condenser, Core Flooding System etc.). On March 21st, 2013, the worldwide first large-scale only passively managed integral accident test of a boiling water reactor was simulated at INKA. The integral test measured the combined response of the KERENA passive safety systems to the postulated initiating event was the “Main Steam Line Break” (MSLB) inside the Containment with decay heat simulation. The results of the performed integral test (MSLB) showed that the passive safety systems alone are capable to bring the plant to stable conditions meeting all required safety targets with sufficient margins. Therefore the test verified the function of those components and the interplay between them as response to an anticipated accident scenario

  14. Philosophy of safety evaluation on fast breeder reactor

    International Nuclear Information System (INIS)

    1981-01-01

    This is the report submitted from the special subcommittee on reactor safety standard to the Nuclear Safety Commission on October 14, 1980, and it was decided to temporarily apply this concept to the safety examination on fast breeder reactors. The examination and discussion of this report were performed by taking the prototype reactor ''Monju'' into consideration, which is to be the present target, referring to the philosophy of the safety evaluation on fast breeder reactors in foreign countries and based on the experiences in the fast experimental reactor ''Joyo''. The items applicable to the safety evaluation for liquid metal-cooled fast breeder reactors (LMFBR) as they are among the existing safety examination guidelines are applied. In addition to the existing guidelines, the report describes the matters to be considered specifically for core, fuel, sodium, sodium void, reactor shut-down system, reactor coolant boundary, cover gas boundary and others, intermediate cooling system, removal of decay heat, containment vessels, high temperature structures, and aseismatic property in the safety design of LMFBR's. For the safety evaluation for LMFBR's, the abnormal transient changes in operation and the phenomena to be evaluated as accidents are enumerated. In order to judge the propriety of the criteria of locating LMFBR facilities, the serious and hypothetical accidents are decided to be evaluated in accordance with the guideline for reactor location investigation. (Wakatsuki, Y.)

  15. Nuclear space power safety and facility guidelines study

    International Nuclear Information System (INIS)

    Mehlman, W.F.

    1995-01-01

    This report addresses safety guidelines for space nuclear reactor power missions and was prepared by The Johns Hopkins University Applied Physics Laboratory (JHU/APL) under a Department of Energy grant, DE-FG01-94NE32180 dated 27 September 1994. This grant was based on a proposal submitted by the JHU/APL in response to an open-quotes Invitation for Proposals Designed to Support Federal Agencies and Commercial Interests in Meeting Special Power and Propulsion Needs for Future Space Missionsclose quotes. The United States has not launched a nuclear reactor since SNAP 10A in April 1965 although many Radioisotope Thermoelectric Generators (RTGs) have been launched. An RTG powered system is planned for launch as part of the Cassini mission to Saturn in 1997. Recently the Ballistic Missile Defense Office (BMDO) sponsored the Nuclear Electric Propulsion Space Test Program (NEPSTP) which was to demonstrate and evaluate the Russian-built TOPAZ II nuclear reactor as a power source in space. As of late 1993 the flight portion of this program was canceled but work to investigate the attributes of the reactor were continued but at a reduced level. While the future of space nuclear power systems is uncertain there are potential space missions which would require space nuclear power systems. The differences between space nuclear power systems and RTG devices are sufficient that safety and facility requirements warrant a review in the context of the unique features of a space nuclear reactor power system

  16. Proceeding of the 7. Seminar on Technology and Safety of Nuclear Power Plants and Nuclear Facilities

    International Nuclear Information System (INIS)

    Hastowo, Hudi; Antariksawan, Anhar R.; Soetrisnanto, Arnold Y; Jujuratisbela, Uju; Aziz, Ferhat; Su'ud, Zaki; Suprawhardana, M. Salman

    2002-02-01

    The seventh proceedings of seminar safety and technology of nuclear power plant and nuclear facilities, held by National Nuclear Energy Agency. The Aims of seminar is to exchange and disseminate information about safety and nuclear Power Plant Technology and Nuclear Facilities consist of technology; high temperature reactor and application for national development sustain able and high technology. This seminar level all aspects technology, Power Reactor research reactor, high temperature reactor and nuclear facilities. The article is separated by index

  17. Fast breeder reactor safety : a perspective

    International Nuclear Information System (INIS)

    Kale, R.D.

    1992-01-01

    Taking into consideration India's limited reserves of natural and vast reserves of thorium, the fast reactor route holds a great promise for India's energy supply in future. The fast reactor fueled with 239 Pu/ 238 U (unused or depleted) produces (breeds) more fissionable fuel material 239 Pu than it consumes. Calculations show that a fast breeder reactor (FBR) increases energy potential of natural uranium by about 60 times. As the fast reactor can also convert 232 Th into 233 U which is a fissionable material, it can make India's thorium reserves a source of almost inexhaustible energy supply for a long time to come. Significant advantage of FBR plants cooled by sodium and their world-wide operating experience are reviewed. There are two main safety issues of FBR, one nuclear and the other non-nuclear. The nuclear issue concerns core disruptive accident and the non-nuclear one concerns the high chemical energy potential of sodium. These two issues are analysed and it is pointed that they are manageable by current design, construction and operational practices. Main findings of safety research during the last six to eight years in West European Countries and United States of America (US) are summarised. Three stage engineered safety provision incorporated into the design of the sodium cooled Fast Breeder Test Reactor (FBTR) commissioned at Kalpakkam are explained. The important design safety features of FBTR such as primary system containment, emergency core cooling, plant protection system, inherent safety features achieved through reactivity coefficients, and natural convection cooling are discussed. Theoretical analysis and experimental research in fast reactor safety carried out at the Indira Gandhi Centre for Atomic Research during the past some years are reviewed. (M.G.B.)

  18. Discussion of the use of the Dragon reactor as a facility for integral reactor physics experiments

    Energy Technology Data Exchange (ETDEWEB)

    Gutmann, H

    1972-06-05

    The purpose and use of the Dragon Reactor Experiment (DRE) has changed considerably during the years of its operation. The original purpose was to show that the principle of a High Temperature Reactor is sound and demonstrate its operation. After this achievement, the purpose of the Dragon reactor changed to the use as a fuel testing facility. During recent years, a new use of the DRE has been added to its use as a fuel testing facility, namely Fuel Element Design Testing. The current report covers reactor physics experiments aspects.

  19. STARFIRE remote maintenance and reactor facility concept

    International Nuclear Information System (INIS)

    Graumann, D.W.; Field, R.E.; Lutz, G.R.; Trachsel, C.A.

    1981-01-01

    A total remote maintenance facility has been designed for all equipment located within the reactor building and hot cell, although operational flexibility has been provided by design of the reactor shielding such that personnel access into the reactor building within 24 hours after reactor shutdown is possible. The reactor design permits removal and replacement of all components if necessary, however, the vacuum pumps, isolation valves and blanket require scheduled, routine maintenance. Reactor scheduled maintenance does not dominate annual plant downtime, therefore, several scheduled operations can be added without affecting reactor availability. The maintenance facilities consist of the reactor building, the hot cell, the reactor service area and the remote maintenance control room. The reactor building contains the reactor, selected support system modules, and required maintenance equipment. The reactor and the support systems are maintained with (1) equipment that is mounted on a monorail system; (2) overhead cranes; and (3) bridge-mounted electromechanical manipulators. The hot cell is located outside of the reactor building to localize contamination products and permit independent operation. An equipment air lock connects the reactor building to the hot cell

  20. Life Management and Safety of Nuclear Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Fabbri, S.; Diluch, A.; Vega, G., E-mail: fabbri@cnea.gov.ar [Comisión Nacional de Energía Atómica, Buenos Aires (Argentina)

    2014-10-15

    The nuclear programme in Argentina includes: nuclear power and related supplies, medical and industrial applications, waste management, research and development and human training. Nuclear facilities require life management programs that allow a safe operation. Safety is the first priority for designers and operators. This can be attained with defence in depth: regular inspections and maintenance procedures to minimize failure risks. CNEA objectives in this area are to possess the necessary capability to give safe and fast technical support. Within this scheme, one of the main activities undertaken by CNEA is to provide technological assistance to the nuclear plants and research reactors. As a consequence of an increasing concern about safety and ageing a Life Management Department for safe operation was created to take care of these subjects. The goal is to elaborate a Safety Evaluation Process for the critical components of nuclear plants and other facilities. The overall objectives of a safety process are to ensure a continuous safe, reliable and effective operation of nuclear facilities and it means the implementation of the defence in deep concept to enhance safety for the protection of the public, the workers and the environment. (author)

  1. Nuclear Safety Research and Facilities Department. Annual report 1999

    Energy Technology Data Exchange (ETDEWEB)

    Majborn, B.; Damkjaer, A.; Hedemann Jensen, P.; Nielsen, S.P.; Nonboel, E. [eds.

    2000-04-01

    The report presents a summary of the work of the Nuclear Safety Research and Facilities Department in 1999. The department's research and development activities were organized in two research programmes: 'Radiation Protection and Reactor Safety' and 'Radioecology and Tracer Studies'. The nuclear facilities operated by the department include the research reactor DR 3, the Isotope Laboratory, the Waste Management Plant, and the educational reactor DR 1. Lists of staff and publications are included together with a summary of the staff's participation in national and international committees. (au)

  2. Nuclear Safety Research and Facilities Department annual report 1999

    DEFF Research Database (Denmark)

    Majborn, B.; Damkjær, A.; Jensen, Per Hedemann

    2000-01-01

    The report presents a summary of the work of the Nuclear Safety Research and Facilities Department in 1999. The department´s research and development activities were organized in two research programmes: "Radiation Protection and Reactor Safety" and"Radioecology and Tracer Studies". The nuclear...... facilities operated by the department include the research reactor DR 3, the Isotope Laboratory, the Waste Management Plant, and the educational reactor DR 1. Lists of staff and publications are includedtogether with a summary of the staff´s participation in national and international committees....

  3. Nuclear Safety Research and Facilities Department annual report 1997

    Energy Technology Data Exchange (ETDEWEB)

    Majborn, B.; Aarkrog, A.; Brodersen, K. [and others

    1998-04-01

    The report presents a summary of the work of the Nuclear Safety Research and Facilities Department in 1997. The department`s research and development activities were organized in four research programmes: Reactor Safety, Radiation protection, Radioecology, and Radioanalytical Chemistry. The nuclear facilities operated by the department include the research reactor DR3, the Isotope Laboratory, the Waste Treatment Plant, and the educational reactor DR1. Lists of staff and publications are included together with a summary of the staff`s participation in national and international committees. (au) 11 tabs., 39 ills.; 74 refs.

  4. Nuclear Safety Research and Facilities Department annual report 1998

    Energy Technology Data Exchange (ETDEWEB)

    Majborn, B.; Brodersen, K.; Damkjaer, A.; Hedemann Jensen, P.; Nielsen, S.P.; Nonboel, E

    1999-04-01

    The report present a summary of the work of the Nuclear Safety Research and Facilities Department in 1998. The department`s research and development activities were organized in two research programmes: `Radiation Protection and Reactor Safety` and `Radioecology and Tracer Studies`. The nuclear facilities operated by the department include the research reactor DR3, the Isotope Laboratory, the Waste Treatment plant, and the educational reactor DR1. Lsits of staff and publications are included together with a summary of the staff`s participation in national and international committees. (au)

  5. Nuclear Safety Research and Facilities Department. Annual report 1999

    International Nuclear Information System (INIS)

    Majborn, B.; Damkjaer, A.; Hedemann Jensen, P.; Nielsen, S.P.; Nonboel, E.

    2000-04-01

    The report presents a summary of the work of the Nuclear Safety Research and Facilities Department in 1999. The department's research and development activities were organized in two research programmes: 'Radiation Protection and Reactor Safety' and 'Radioecology and Tracer Studies'. The nuclear facilities operated by the department include the research reactor DR 3, the Isotope Laboratory, the Waste Management Plant, and the educational reactor DR 1. Lists of staff and publications are included together with a summary of the staff's participation in national and international committees. (au)

  6. Nuclear Safety Research and Facilities department annual report 1996

    International Nuclear Information System (INIS)

    Majborn, B.; Brodersen, K.; Damkjaer, A.; Floto, H.; Heydorn, K.; Oelgaard, P.L.

    1997-04-01

    The report presents a summary of the work of the Nuclear Safety Research and Facilities Department in 1996. The Department's research and development activities are organized in three research programmes: Radiation Protection, Reactor Safety, and Radioanalytical Chemistry. The nuclear facilities operated by the department include the Research Reactor DR3, the Isotope Laboratory, the Waste Treatment Plant, and the Educational Reactor DR1. Lists of staff and publications are included together with a summary of the staff's participation in national and international committees. (au) 2 tabs., 28 ills

  7. Nuclear Safety Research and Facilities Department annual report 1997

    International Nuclear Information System (INIS)

    Majborn, B.; Aarkrog, A.; Brodersen, K.

    1998-04-01

    The report presents a summary of the work of the Nuclear Safety Research and Facilities Department in 1997. The department's research and development activities were organized in four research programmes: Reactor Safety, Radiation protection, Radioecology, and Radioanalytical Chemistry. The nuclear facilities operated by the department include the research reactor DR3, the Isotope Laboratory, the Waste Treatment Plant, and the educational reactor DR1. Lists of staff and publications are included together with a summary of the staff's participation in national and international committees. (au)

  8. Nuclear Safety Research and Facilities Department annual report 1998

    International Nuclear Information System (INIS)

    Majborn, B.; Brodersen, K.; Damkjaer, A.; Hedemann Jensen, P.; Nielsen, S.P.; Nonboel, E.

    1999-04-01

    The report present a summary of the work of the Nuclear Safety Research and Facilities Department in 1998. The department's research and development activities were organized in two research programmes: 'Radiation Protection and Reactor Safety' and 'Radioecology and Tracer Studies'. The nuclear facilities operated by the department include the research reactor DR3, the Isotope Laboratory, the Waste Treatment plant, and the educational reactor DR1. Lsits of staff and publications are included together with a summary of the staff's participation in national and international committees. (au)

  9. Experimental facilities for Generation IV reactors research

    International Nuclear Information System (INIS)

    Krecanova, E.; Di Gabriele, F.; Berka, J.; Zychova, M.; Macak, J.; Vojacek, A.

    2013-06-01

    Centrum Vyzkumu Rez (CVR) is research and development Company situated in Czech Republic and member of the UJV group. One of its major fields is material research for Generation IV reactor concepts, especially supercritical water-cooled reactor (SCWR), very high temperature/gas-cooled fast reactor (VHTR/GFR) and lead-cooled fast reactor (LFR). The CVR is equipped by and is building unique experimental facilities which simulate the environment in the active zones of these reactor concepts and enable to pre-qualify and to select proper constructional materials for the most stressed components of the facility (cladding, vessel, piping). New infrastructure is founded within the Sustainable Energy project focused on implementation the Generation IV and fusion experimental facilities. The research of SCWR concept is divided to research and development of the constructional materials ensured by SuperCritical Water Loop (SCWL) and fuel components research on Fuel Qualification Test loop (SCWL-FQT). SCWL provides environment of the primary circuits of European SCWR, pressure 25 MPa, temperature 600 deg. C and its major purpose is to simulate behavior of the primary medium and candidate constructional materials. On-line monitoring system is included to collect the operational data relevant to experiment and its evaluation (pH, conductivity, chemical species concentration). SCWL-FQT is facility focused on the behavior of cladding material and fuel at the conditions of so-called preheater, the first pass of the medium through the fuel (in case of European SCWR concept). The conditions are 450 deg. C and 25 MPa. SCWL-FQT is unique facility enabling research of the shortened fuel rods. VHTR/GFR research covers material testing and also cleaning methods of the medium in primary circuit. The High Temperature Helium Loop (HTHL) enables exposure of materials and simulates the VHTR/GFR core environment to analyze the behavior of medium, especially in presence of organic compounds and

  10. Unusual occurrences in fast breeder test reactor

    International Nuclear Information System (INIS)

    Kapoor, R.P.; Srinivasan, G.; Ellappan, T.R.; Ramalingam, P.V.; Vasudevan, A.T.; Iyer, M.A.K.; Lee, S.M.; Bhoje, S.B.

    2000-01-01

    Fast Breeder Test Reactor (FBTR) is a 40 MWt/13.2 MWe sodium cooled mixed carbide fuelled reactor. Its main aim is to generate experience in the design, construction and operation of fast reactors including sodium systems and to serve as an irradiation facility for the development of fuel and structural materials for future fast reactors. It achieved first criticality in Oct 85 with Mark I core (70% PuC - 30% UC). Steam generator was put in service in Jan 93 and power was raised to 10.5 MWt in Dec 93. Turbine generator was synchronised to the grid in Jul 97. The indigenously developed mixed carbide fuel has achieved a burnup of 44,000 MW-d/t max at a linear heat rating of 320 W/cm max without any fuel clad failure. The commissioning and operation of sodium systems and components have been smooth and performance of major components, viz., sodium pumps, intermediate heat exchangers and once through sodium heated steam generators (SG) have been excellent. There have been three minor incidents of Na/NaK leaks during the past 14 years, which are described in the paper. There have been no incident of a tube leak in SG. However, three incidents of water leaks from water / steam headers have been detailed. The plant has encountered some unusual occurrences, which were critically analysed and remedial measures, in terms of system and procedural modifications, incorporated to prevent recurrence. This paper describes unusual occurrences of fuel handling incident of May 1987, main boiler feed pump seizure in Apr 1992, reactivity transients in Nov 1994 and Apr 1995, and malfunctioning of the core cover plate mechanism in Jul 1995. These incidents have resulted in long plant shutdowns. During the course of investigation, various theoretical and experimental studies were carried out for better understanding of the phenomena and several inspection techniques and tools were developed resulting in enriching the technology of sodium cooled reactors. FBTR has 36 neutronic and process

  11. Safety philosophy and safety technology of the Soviet RBMK reactors

    International Nuclear Information System (INIS)

    Zuend, H.; Jarvis, A.S.; Haennis, H.P.; Tikal, J.

    1986-01-01

    Safety requirements and control in USSR are outlined. Safety criteria and practical application in the case of the RBMK type reactor Chernobyl-4 are discussed. An overview of the Chernobyl-4 reactor accident including its causes is given. Measures to improve the safety of RBMK reactors are described

  12. ORNL results for Test Case 1 of the International Atomic Energy Agency's research program on the safety assessment of Near-Surface Radioactive Waste Disposal Facilities

    International Nuclear Information System (INIS)

    Thorne, D.J.; McDowell-Boyer, L.M.; Kocher, D.C.; Little, C.A.; Roemer, E.K.

    1993-01-01

    The International Atomic Energy Agency (IAEA) started the Coordinated Research Program entitled '''The Safety Assessment of Near-Surface Radioactive Waste Disposal Facilities.'' The program is aimed at improving the confidence in the modeling results for safety assessments of waste disposal facilities. The program has been given the acronym NSARS (Near-Surface Radioactive Waste Disposal Safety Assessment Reliability Study) for ease of reference. The purpose of this report is to present the ORNL modeling results for the first test case (i.e., Test Case 1) of the IAEA NSARS program. Test Case 1 is based on near-surface disposal of radionuclides that are subsequently leached to a saturated-sand aquifer. Exposure to radionuclides results from use of a well screened in the aquifer and from intrusion into the repository. Two repository concepts were defined in Test Case 1: a simple earth trench and an engineered vault

  13. Magnox Electric Littlebrook reactor inspection and repair rehearsal facility

    International Nuclear Information System (INIS)

    Barnes, S.A.; Clayton, R.; Gaydon, B.G.; Ramsey, B.H.

    1996-01-01

    Magnox reactors, although designed to be maintenance free during their operational life, have nevertheless highlighted the need for test rig facilities to train operators in the methods and techniques of reactor inspection and repair. The history of the facility for reactor engineering development (FRED) is described and its present role as a repair rehearsal facility noted. Advances in computer graphics may, in future, mean that such operator training will be virtual reality rather than analog reality based; however the need for such rigs to commission techniques and equipment and to establish performance and reliability is likely to continue. (UK)

  14. Decommissioning of nuclear fuel cycle facilities. Safety guide

    International Nuclear Information System (INIS)

    2001-01-01

    The objective of this Safety Guide is to provide guidance to regulatory bodies and operating organizations on planning and provision for the safe management of the decommissioning of non-reactor nuclear fuel cycle facilities. While the basic safety considerations for the decommissioning of nuclear fuel cycle facilities are similar to those for nuclear power plants, there are important differences, notably in the design and operating parameters for the facilities, the type of radioactive material and the support systems available. It is the objective of this Safety Guide to provide guidance for the shutdown and eventual decommissioning of such facilities, their individual characteristics being taken into account

  15. Testing experience with fast flux test facility

    International Nuclear Information System (INIS)

    Noordhoff, B.H.; McGough, C.B.; Nolan, J.E.

    1975-01-01

    Early FFTF project planning emphasized partial and full-scale testing of major reactor and plant prototype components under expected environmental conditions, excluding radiation fields. Confirmation of component performance during FFTF service was considered essential before actual FFTF startup, to provide increased assurance against FFTF startup delays or operational difficulties and downtime. Several new sodium facilities were constructed, and confirmation tests on the prototype components are now in progress. Test conditions and results to date are reported for the primary pump, intermediate heat exchanger, sodium-to-air dump heat exchanger, large and small sodium valves, purification cold trap, in-vessel handling machine, instrument tree, core restraint, control rod system, low-level flux monitor, closed loop ex-vessel machine, refueling equipment, and selected maintenance equipment. The significance and contribution of these tests to the FFTF and Liquid Metal Fast Breeder Reactor (LMFBR) program are summarized. (U.S.)

  16. A plan for safety and integrity of research reactor components

    International Nuclear Information System (INIS)

    Moatty, Mona S. Abdel; Khattab, M.S.

    2013-01-01

    Highlights: ► A plan for in-service inspection of research reactor components is put. ► Section XI of the ASME Code requirements is applied. ► Components subjected to inspection and their classes are defined. ► Flaw evaluation and its acceptance–rejection criteria are reviewed. ► A plan of repair or replacement is prepared. -- Abstract: Safety and integrity of a research reactor that has been operated over 40 years requires frequent and thorough inspection of all the safety-related components of the facility. The need of increasing the safety is the need of improving the reliability of its systems. Diligent and extensive planning of in-service inspection (ISI) of all reactor components has been imposed for satisfying the most stringent safety requirements. The Safeguards Officer's responsibilities of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code ASME Code have been applied. These represent the most extensive and time-consuming part of ISI program, and identify the components subjected to inspection and testing, methods of component classification, inspection and testing techniques, acceptance/rejection criteria, and the responsibilities. The paper focuses on ISI planning requirements for welded systems such as vessels, piping, valve bodies, pump casings, and control rod-housing parts. The weld in integral attachments for piping, pumps, and valves are considered too. These are taken in consideration of safety class (1, 2, 3, etc.), reactor age, and weld type. The parts involve in the frequency of inspection, the examination requirements for each inspection, the examination method are included. Moreover the flaw evaluation, the plan of repair or replacement, and the qualification of nondestructive examination personnel are considered

  17. An automated test facility for neutronic amplifiers

    International Nuclear Information System (INIS)

    Beattie, W.J.

    1997-01-01

    Neutronic amplifiers are used at the Chalk River Laboratory in applications such as neutron flux monitoring and reactor control systems. Routine preventive maintenance of control and safety systems included annual calibration and characterization of the neutronic amplifiers. An investigation into the traditional methods of annual routine maintenance of amplifiers concluded that frequency and phase response measurements in particular were labour intensive and subject to non-repeatable errors. A decision was made to upgrade testing methods and facilities by using programmable test equipment under the control of a computer. In order to verify the results of the routine measurements, expressions for the transfer functions were derived from the circuit diagrams. Frequency and phase responses were then calculated and plotted thus providing a bench-mark to which the test results can be compared. (author)

  18. An integral effect test facility of the SMART, SMART ITL

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Sik; Moon, Sang Ki; Kim, Yeon Sik; Cho, Seok; Choi, Ki Yong; Bae, Hwang; Kim, Dong Eok; Choi, Nam Hyun; Min, Kyoung Ho; Ko, Yung Joo; Shin, Yong Cheol; Park, Rae Joon; Lee, Won Jae; Song, Chul Hwa; Yi, Sung Jae [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    SMART (System integrated Modular Advanced ReacTor) is a 330 MWth integral pressurized water reactor (iPWR) developed by KAERI and had obtained standard design approval (SDA) from Korean regulatory authority on July 2012. In this SMART design main components including a pressurizer, reactor coolant pumps and steam generators are installed in a reactor pressure vessel without any large connecting pipes. As the LBLOCA scenario is inherently excluded, its safety systems could be simplified only to ensure the safety during the SBLOCA scenarios and the other system transients. An integral effect test loop for the SMART (SMART ITL), or called as FESTA, had been designed to simulate the integral thermal hydraulic behavior of the SMART. The objectives of the SMART ITL are to investigate and understand the integral performance of reactor systems and components and the thermal hydraulic phenomena occurred in the system during normal, abnormal and emergency conditions, and to verify the system safety during various design basis events of the SMART. The integral effect test data will also be used to validate the related thermal hydraulic models of the safety analysis code such as TASS/SMR S, which is used for performance and accident analysis of the SMART design. This paper introduces the scaling analysis and scientific design of the integral test facility of the SMART, SMART ITL and its scaling analysis results.

  19. Safety issues at the defense production reactors

    International Nuclear Information System (INIS)

    1987-01-01

    The United States produces plutonium and tritium for use in nuclear weapons at the defense production reactors - the N Reactor in Washington and the Savannah River reactors in South Carolina. This report reaches general conclusions about the management of those reactors and highlights a number of safety and technical issues that should be resolved. The report provides an assessment of the safety management, safety review, and safety methodology employed by the Department of Energy and the private contractors who operate the reactors for the federal government. This report examines the safety objective established by the Department of Energy for the production reactors and the process the Department of its contractors use to implement the objective; focuses on a variety of uncertainties concerning the production reactors, particularly those related to potential vulnerabilities to severe accidents; and identifies ways in which the DOE approach to management of the safety of the production reactors can be improved

  20. BWR Full Integral Simulation Test (FIST) program: facility description report

    International Nuclear Information System (INIS)

    Stephens, A.G.

    1984-09-01

    A new boiling water reactor safety test facility (FIST, Full Integral Simulation Test) is described. It will be used to investigate small breaks and operational transients and to tie results from such tests to earlier large-break test results determined in the TLTA. The new facility's full height and prototypical components constitute a major scaling improvement over earlier test facilities. A heated feedwater system, permitting steady-state operation, and a large increase in the number of measurements are other significant improvements. The program background is outlined and program objectives defined. The design basis is presented together with a detailed, complete description of the facility and measurements to be made. An extensive component scaling analysis and prediction of performance are presented

  1. Advanced reactor concepts and safety

    International Nuclear Information System (INIS)

    Lipsett, J.J.

    1988-06-01

    The need for some consistency in the terms used to describe the evolution of methods for ensuring the safety of nuclear reactors has been identified by the IAEA. This is timely since there appears to be a danger that the precision of many valuable words is being diluted and that a new jargon may appear that will confuse rather than aid the communication of important but possibly diverse philosophies and concepts. Among the difficulties faced by the nuclear industry is promoting and gaining a widespread understanding of the risks actually posed by nuclear reactors. In view of the importance of communication to both the public and to the technical community generally, the starting point for the definition of terms must be with dictionary meanings and common technical usage. The nuclear engineering community should use such words in conformance with the whole technical world. This paper addresses many of the issues suggested in the invitation to meet and also poses some additional issues for consideration. Some examples are the role of the operator in either enhancing or degrading safety and how the meaning or interpretation of the word 'safety' can be expected to change during the next few decades. It is advantageous to use criteria against which technologies and ongoing operating performance can be judged provided that the criteria are generic and not specific to particular reactor concepts. Some thoughts are offered on the need to frame the criteria carefully so that innovative solutions and concepts are fostered, not stifled

  2. Helium heater design for the helium direct cycle component test facility. [for gas-cooled nuclear reactor power plant

    Science.gov (United States)

    Larson, V. R.; Gunn, S. V.; Lee, J. C.

    1975-01-01

    The paper describes a helium heater to be used to conduct non-nuclear demonstration tests of the complete power conversion loop for a direct-cycle gas-cooled nuclear reactor power plant. Requirements for the heater include: heating the helium to a 1500 F temperature, operating at a 1000 psia helium pressure, providing a thermal response capability and helium volume similar to that of the nuclear reactor, and a total heater system helium pressure drop of not more than 15 psi. The unique compact heater system design proposed consists of 18 heater modules; air preheaters, compressors, and compressor drive systems; an integral control system; piping; and auxiliary equipment. The heater modules incorporate the dual-concentric-tube 'Variflux' heat exchanger design which provides a controlled heat flux along the entire length of the tube element. The heater design as proposed will meet all system requirements. The heater uses pressurized combustion (50 psia) to provide intensive heat transfer, and to minimize furnace volume and heat storage mass.

  3. Climatic Environmental Test Facilities

    Data.gov (United States)

    Federal Laboratory Consortium — RTTC has an extensive suite of facilities for supporting MIL-STD-810 testing, toinclude: Temperature/Altitude, Rapid Decompression, Low/High Temperature,Temperature...

  4. Wind Tunnel Testing Facilities

    Data.gov (United States)

    Federal Laboratory Consortium — NASA Ames Research Center is pleased to offer the services of our premier wind tunnel facilities that have a broad range of proven testing capabilities to customers...

  5. Nuclear materials facility safety initiative

    International Nuclear Information System (INIS)

    Peddicord, K.L.; Nelson, P.; Roundhill, M.; Jardine, L.J.; Lazarev, L.; Moshkov, M.; Khromov, V.V.; Kruchkov, E.; Bolyatko, V.; Kazanskij, Yu.; Vorobeva, I.; Lash, T.R.; Newton, D.; Harris, B.

    2000-01-01

    Safety in any facility in the nuclear fuel cycle is a fundamental goal. However, it is recognized that, for example, should an accident occur in either the U.S. or Russia, the results could seriously delay joint activities to store and disposition weapons fissile materials in both countries. To address this, plans are underway jointly to develop a nuclear materials facility safety initiative. The focus of the initiative would be to share expertise which would lead in improvements in safety and safe practices in the nuclear fuel cycle.The program has two components. The first is a lab-to-lab initiative. The second involves university-to-university collaboration.The lab-to-lab and university-to-university programs will contribute to increased safety in facilities dealing with nuclear materials and related processes. These programs will support important bilateral initiatives, develop the next generation of scientists and engineers which will deal with these challenges, and foster the development of a safety culture

  6. Complementary Safety Assessments for Research Reactors for the French Nuclear Safety Authority

    International Nuclear Information System (INIS)

    Kassiotis, Christophe; Rigaud, Antoine; Evrard, Lydie

    2013-01-01

    The 'Autorite de surete nucleaire' (ASN) requested licensees to undertake stress tests, called complementary safety assessments (CSA), of their installations on May 5th 2011, following the accident that occurred in Japan on March 11th 2011. Their mission consisted in providing feedback on the consequences of potential extreme events. In this process, all the French facilities were divided into three categories of decreasing priority, depending on two main factors: on the one hand, their vulnerability to the various phenomena that led to the Fukushima accident, and on the other hand, the amount of radioactive elements that would be dispersed in the event of a failure of the safety functions. On the 79 high-priority facilities, only five of them are research or experimental reactors (including two currently shutdown or in decommissioning) and their operators (the 'Comissariat a l'energie atomique et aux energies alternatives' (CEA) and the 'Institut Laue Langevin') submitted their reports to the ASN on September 15 th 2011. Concerning the lower-priority facilities, including three other facilities (two research reactors operated by the CEA and a facility operated by ITER Organization) the deadline was September 15 th 2012. Finally, the remaining facilities were not asked to submit a report yet, but they will have to do it later, mainly on the occasion of their next periodic safety review. The analyses of the cliff-edge effects, that may occur in extreme situations (exceptional scale event, combination of several disasters...), led to the definition of a hardened safety core concept by the 'Institut de radioprotection et de surete nucleaire' (IRSN). This hardened safety core of structures, equipment and organizational measures must ensure the ultimate protection of the concerned facilities in extreme situations : it is designed to prevent severe accidents (or curb their progression), limit large scale releases for extreme accidents, and enables the operating teams to

  7. Methods and strategies for future reactor safety goals

    Science.gov (United States)

    Arndt, Steven Andrew

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

  8. Proceedings of the international topical meeting on advanced reactors safety: Volume 2

    International Nuclear Information System (INIS)

    1997-01-01

    In this volume, 89 papers are grouped under the following headings: advances in research/test reactor safety; advanced reactor accident management and emergency actions; advanced reactors instrumentation/controls/human factors; probabilistic risk/safety and reliability assessments; steam explosion research and issues; advanced reactor severe accident issues and research (analysis and assessments); advanced reactor thermal hydraulics; accelerator-driven source safety; liquid-metal reactor safety; structural assessments and issues; late papers

  9. Safety culture in a major nuclear fuel cycle facility

    International Nuclear Information System (INIS)

    Pushparaja; Abani, M.C.

    2002-01-01

    Human factor plays an important role in development of safety culture in any nuclear fuel cycle facility. This is more relevant in major nuclear facility such as a reactor or a reprocessing plant. In Indian reprocessing plants, an effective worker's training, education and certification program is in place to sensitize the worker's response to safety and safe work procedures. The methodology followed to self evaluation of safety culture and the benefits in a reprocessing plant is briefly discussed. Various indicators of safety performance and visible signs of a good safety management are also qualitatively analyzed. (author)

  10. Passive BWR integral LOCA testing at the Karlstein test facility INKA

    Energy Technology Data Exchange (ETDEWEB)

    Drescher, Robert [AREVA GmbH, Erlangen (Germany); Wagner, Thomas [AREVA GmbH, Karlstein am Main (Germany); Leyer, Stephan [TH University of Applied Sciences, Deggendorf (Germany)

    2014-05-15

    KERENA is an innovative AREVA GmbH boiling water reactor (BWR) with passive safety systems (Generation III+). In order to verify the functionality of the reactor design an experimental validation program was executed. Therefore the INKA (Integral Teststand Karlstein) test facility was designed and erected. It is a mockup of the BWR containment, with integrated pressure suppression system. While the scaling of the passive components and the levels match the original values, the volume scaling of the containment compartments is approximately 1:24. The storage capacity of the test facility pressure vessel corresponds to approximately 1/6 of the KERENA RPV and is supplied by a benson boiler with a thermal power of 22 MW. In March 2013 the first integral test - Main Steam Line Break (MSLB) - was executed. The test measured the combined response of the passive safety systems to the postulated initiating event. The main goal was to demonstrate the ability of the passive systems to ensure core coverage, decay heat removal and to maintain the containment within defined limits. The results of the test showed that the passive safety systems are capable to bring the plant to stable conditions meeting all required safety targets with sufficient margins. Therefore the test verified the function of those components and the interplay between them. The test proved that INKA is an unique test facility, capable to perform integral tests of passive safety concepts under plant-like conditions. (orig.)

  11. Preventive and Predictive Maintenance, Warehousing of Spares, Periodic Testing and In-Service Inspection Activities at the Nigerian Research Reactor-1 Facility

    Energy Technology Data Exchange (ETDEWEB)

    Yusuf, I.; Mati, A. A.; Dewu, B. B.M., [Centre for Energy Research and Training, Ahmadu Bello University, Zaria (Nigeria)

    2014-08-15

    The Nigerian Research Reactor–1, or NIRR-1, is sited at Centre for Energy Research and Training, Ahmadu Bello University, Zaria, Nigeria. Activities on preventive or routine maintenance have been institutionalized since the commissioning of the reactor in February 2004. This has grossly reduced the rates of corrective maintenance activities and helped the reactor management a great deal in predicting failure rates of reactor components and other auxiliary units. Routine maintenance of systems and components are being carried out on a weekly, quarterly and annual basis based on manufacturer’s recommendations, which have been reviewed and improved over the years. The paper presents the implementation of maintenance activities in NIRR-1 from its initial criticality in 2004 till today and the new scheme for periodic testing and in-service-inspection developed after an IAEA Integrated Safety Assessment of Research Reactors mission. The measures put in place are envisaged to reduce the negative impact of ageing on NIRR-1 and its auxiliary systems. (author)

  12. Toroid magnet test facility

    CERN Multimedia

    2002-01-01

    Because of its exceptional size, it was not feasible to assemble and test the Barrel Toroid - made of eight coils - as an integrated toroid on the surface, prior to its final installation underground in LHC interaction point 1. It was therefore decided to test these eight coils individually in a dedicated test facility.

  13. Test and User Facilities | NREL

    Science.gov (United States)

    Test and User Facilities Test and User Facilities Our test and user facilities are available to | L | M | N | O | P | Q | R | S | T | U | V | W | X | Y | Z B Battery Thermal and Life Test Facility Biochemical Conversion Pilot Plant C Controllable Grid Interface Test System D Dynamometer Test Facilities

  14. Updates to the Generation of Physics Data Inputs for MAMMOTH Simulations of the Transient Reactor Test Facility - FY2016

    Energy Technology Data Exchange (ETDEWEB)

    Ortensi, Javier [Idaho National Lab. (INL), Idaho Falls, ID (United States); Baker, Benjamin Allen [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schunert, Sebastian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wang, Yaqi [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gleicher, Frederick Nathan [Idaho National Lab. (INL), Idaho Falls, ID (United States); DeHart, Mark David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-06-01

    The INL is currently evolving the modeling and simulation (M&S) capability that will enable improved core operation as well as design and analysis of TREAT experiments. This M&S capability primarily uses MAMMOTH, a reactor physics application being developed under Multi-physics Object Oriented Simulation Environment (MOOSE) framework. MAMMOTH allows the coupling of a number of other MOOSE-based applications. This second year of work has been devoted to the generation of a deterministic reference solution for the full core, the preparation of anisotropic diffusion coefficients, the testing of the SPH equivalence method, and the improvement of the control rod modeling. In addition, this report includes the progress made in the modeling of the M8 core configuration and experiment vehicle since January of this year.

  15. Experimental Facilities for Performance Evaluation of Fast Reactor Components

    International Nuclear Information System (INIS)

    Chandramouli, S.; Kumar, V.A. Suresh; Shanmugavel, M.; Vijayakumar, G.; Vinod, V.; Noushad, I.B.; Babu, B.; Kumar, G. Padma; Nashine, B.K.; Rajan, K.K.

    2013-01-01

    Brief details about various experimental facilities catering to the testing and performance evaluation requirements of fast reactor components have been brought out. These facilities have been found to be immensely useful to continue research and development activities in the areas of component development and testing, sodium technology, thermal hydraulics and sodium instrumentation for the SFR’s. In addition new facilities which have been planned will be of great importance for the developmental activities related to future SFR’s

  16. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Baek, Won Pil; Song, C. H.; Kim, Y. S.

    2007-02-01

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform the tests for design, operation, and safety regulation of pressurized water reactors. In the first phase of this project (1997.8∼2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished. In the second phase (2002.4∼2005.2), an optimized design of the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) was established and the construction of the facility was almost completed. In the third phase (2005.3∼2007.2), the construction and commission tests of the ATLAS are to be completed and some first-phase tests are to be conducted

  17. Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Perry, E.; Chrzanowski, J.; Rule, K.; Viola, M.; Williams, M.; Strykowsky, R.

    1999-01-01

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. The Decontamination and Decommissioning (D and D) of the TFTR is scheduled to occur over a period of three years beginning in October 1999. This is not a typical Department of Energy D and D Project where a facility is isolated and cleaned up by ''bulldozing'' all facility and hardware systems to a greenfield condition. The mission of TFTR D and D is to: (a) surgically remove items which can be re-used within the DOE complex, (b) remove tritium contaminated and activated systems for disposal, (c) clear the test cell of hardware for future reuse, (d) reclassify the D-site complex as a non-nuclear facility as defined in DOE Order 420.1 (Facility Safety) and (e) provide data on the D and D of a large magnetic fusion facility. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The record-breaking deuterium-tritium experiments performed on TFTR resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 Mev neutrons. The total tritium content within the vessel is in excess of 7,000 Curies while dose rates approach 75 mRem/hr. These radiological hazards along with the size and shape of the Tokamak present a unique and challenging task for dismantling

  18. Oak Ridge rf Test Facility

    International Nuclear Information System (INIS)

    Gardner, W.L.; Hoffman, D.J.; McCurdy, H.C.; McManamy, T.J.; Moeller, J.A.; Ryan, P.M.

    1985-01-01

    The rf Test Facility (RFTF) of Oak Ridge National Laboratory (ORNL) provides a national facility for the testing and evaluation of steady-state, high-power (approx.1.0-MW) ion cyclotron resonance heating (ICRH) systems and components. The facility consists of a vacuum vessel and two fully tested superconducting development magnets from the ELMO Bumpy Torus Proof-of-Principle (EBT-P) program. These are arranged as a simple mirror with a mirror ratio of 4.8. The axial centerline distance between magnet throat centers is 112 cm. The vacuum vessel cavity has a large port (74 by 163 cm) and a test volume adequate for testing prototypic launchers for Doublet III-D (DIII-D), Tore Supra, and the Tokamak Fusion Test Reactor (TFTR). Attached to the internal vessel walls are water-cooled panels for removing the injected rf power. The magnets are capable of generating a steady-state field of approx.3 T on axis in the magnet throats. Steady-state plasmas are generated in the facility by cyclotron resonance breakdown using a dedicated 200-kW, 28-GHz gyrotron. Available rf sources cover a frequency range of 2 to 200 MHz at 1.5 kW and 3 to 18 MHz at 200 kW, with several sources at intermediate parameters. Available in July 1986 will be a >1.0-MW, cw source spanning 40 to 80 MHz. 5 figs

  19. 340 waste handling facility interim safety basis

    Energy Technology Data Exchange (ETDEWEB)

    VAIL, T.S.

    1999-04-01

    This document presents an interim safety basis for the 340 Waste Handling Facility classifying the 340 Facility as a Hazard Category 3 facility. The hazard analysis quantifies the operating safety envelop for this facility and demonstrates that the facility can be operated without a significant threat to onsite or offsite people.

  20. 340 waste handling facility interim safety basis

    International Nuclear Information System (INIS)

    VAIL, T.S.

    1999-01-01

    This document presents an interim safety basis for the 340 Waste Handling Facility classifying the 340 Facility as a Hazard Category 3 facility. The hazard analysis quantifies the operating safety envelop for this facility and demonstrates that the facility can be operated without a significant threat to onsite or offsite people