WorldWideScience

Sample records for safety test facility

  1. 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

  2. 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

  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. 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.

  5. 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

  6. 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

  7. 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

  8. 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.

  9. 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

  10. 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.

  11. 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.

  12. 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

  13. 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

  14. 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)

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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.)

  5. 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

  6. 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

  7. 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)

  8. 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

  9. 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

  10. 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)

  11. 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

  12. 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)

  13. 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)

  14. 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.)

  15. 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)

  16. 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

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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

  2. 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

  3. 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.

  4. 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...

  5. 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...

  6. 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

  7. 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.

  8. 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)

  9. 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...

  10. 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...

  11. 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

  12. 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.

  13. 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...

  14. 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.

  15. 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

  16. 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...

  17. 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...

  18. 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...

  19. 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.

  20. 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.

  1. 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)

  2. 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...

  3. 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...

  4. 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...

  5. 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

  6. 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.

  7. 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...

  8. 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...

  9. 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).

  10. 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

  11. 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.

  12. 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

  13. 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.

  14. 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

  15. 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)

  16. 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.)

  17. 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

  18. Distributed Energy Resources Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — NREL's Distributed Energy Resources Test Facility (DERTF) is a working laboratory for interconnection and systems integration testing. This state-of-the-art facility...

  19. Safety overview of the National Ignition Facility

    International Nuclear Information System (INIS)

    Brereton, S.J.; McLouth, L.; Odell, B.; Singh, M.; Tobin, M.; Trent, M.

    1996-01-01

    The National Ignition Facility (NIF) is a proposed US Department of Energy inertial confinement laser fusion facility. The candidate sites for locating the NIF are: Los Alamos National Laboratory, Sandia National Laboratory, the Nevada Test Site, and Lawrence Livermore National Laboratory (LLNL), the preferred site. The NIF will operate by focusing 192 laser beams onto a tiny deuterium- tritium target located at the center of a spherical target chamber. The NIF mission is to achieve inertial confinement fusion (ICF) ignition, access physical conditions in matter of interest to nuclear weapons physics, provide an above ground simulation capability for nuclear weapons effects testing, and contribute to the development of inertial fusion for electrical power production. The NIF has been classified as a radiological, low hazard facility on the basis of a preliminary hazards analysis and according to the DOE methodology for facility classification. This requires that a safety analysis be prepared under DOE Order 5481.1B, Safety Analysis and Review System. A draft Preliminary Safety Analysis Report (PSAR) has been written, and this will be finalized later in 1996. This paper summarizes the safety issues associated with the operation of the NIF. It provides an overview of the hazards, estimates maximum routine and accidental exposures for the preferred site of LLNL, and concludes that the risks from NIF operations are low

  20. 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

  1. Large coil test facility

    International Nuclear Information System (INIS)

    Nelms, L.W.; Thompson, P.B.

    1980-01-01

    Final design of the facility is nearing completion, and 20% of the construction has been accomplished. A large vacuum chamber, houses the test assembly which is coupled to appropriate cryogenic, electrical, instrumentation, diagnostc systems. Adequate assembly/disassembly areas, shop space, test control center, offices, and test support laboratories are located in the same building. Assembly and installation operations are accomplished with an overhead crane. The major subsystems are the vacuum system, the test stand assembly, the cryogenic system, the experimental electric power system, the instrumentation and control system, and the data aquisition system

  2. 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.)

  3. Safety of nuclear fuel cycle facilities. Safety requirements

    International Nuclear Information System (INIS)

    2008-01-01

    This publication covers the broad scope of requirements for fuel cycle facilities that, in light of the experience and present state of technology, must be satisfied to ensure safety for the lifetime of the facility. Topics of specific reference include aspects of nuclear fuel generation, storage, reprocessing and disposal. Contents: 1. Introduction; 2. The safety objective, concepts and safety principles; 3. Legal framework and regulatory supervision; 4. The management system and verification of safety; 5. Siting of the facility; 6. Design of the facility; 7. Construction of the facility; 8. Commissioning of the facility; 9. Operation of the facility; 10. Decommissioning of the facility; Appendix I: Requirements specific to uranium fuel fabrication facilities; Appendix II: Requirements specific to mixed oxide fuel fabrication facilities; Appendix III: Requirements specific to conversion facilities and enrichment facilities

  4. National Solar Thermal Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The National Solar Thermal Test Facility (NSTTF) is the only test facility in the United States of its type. This unique facility provides experimental engineering...

  5. Importance of tests in nuclear facilities

    International Nuclear Information System (INIS)

    Guillemard, B.

    1985-10-01

    In nuclear facilities, safety related systems and equipments are subject, along their whole service-life, to numerous tests. This paper analyses the role of tests in the successive stages of design, construction, exploitation of a nuclear facility. It examines several aspects of test quality control: definition of needs, test planning, intrinsic quality of each test, control of interfaces (test are both the end and the starting point of many actions concerned by quality) and the application [fr

  6. FOOD SAFETY TESTING LABORATORY

    Data.gov (United States)

    Federal Laboratory Consortium — This laboratory develops screening assays, tests and modifies biosensor equipment, and optimizes food safety testing protocols for the military and civilian sector...

  7. Hot Hydrogen Test Facility

    International Nuclear Information System (INIS)

    W. David Swank

    2007-01-01

    The core in a nuclear thermal rocket will operate at high temperatures and in hydrogen. One of the important parameters in evaluating the performance of a nuclear thermal rocket is specific impulse, ISp. This quantity is proportional to the square root of the propellant's absolute temperature and inversely proportional to square root of its molecular weight. Therefore, high temperature hydrogen is a favored propellant of nuclear thermal rocket designers. Previous work has shown that one of the life-limiting phenomena for thermal rocket nuclear cores is mass loss of fuel to flowing hydrogen at high temperatures. The hot hydrogen test facility located at the Idaho National Lab (INL) is designed to test suitability of different core materials in 2500 C hydrogen flowing at 1500 liters per minute. The facility is intended to test non-uranium containing materials and therefore is particularly suited for testing potential cladding and coating materials. In this first installment the facility is described. Automated Data acquisition, flow and temperature control, vessel compatibility with various core geometries and overall capabilities are discussed

  8. 340 Waste Handling Facility interim safety basis

    International Nuclear Information System (INIS)

    Bendixsen, R.B.

    1995-01-01

    This document establishes the interim safety basis (ISB) for the 340 Waste Handling Facility (340 Facility). An ISB is a documented safety basis that provides a justification for the continued operation of the facility until an upgraded final safety analysis report is prepared that complies with US Department of Energy (DOE) Order 5480.23, Nuclear Safety Analysis Reports. The ISB for the 340 Facility documents the current design and operation of the facility. The 340 Facility ISB (ISB-003) is based on a facility walkdown and review of the design and operation of the facility, as described in the existing safety documentation. The safety documents reviewed, to develop ISB-003, include the following: OSD-SW-153-0001, Operating Specification Document for the 340 Waste Handling Facility (WHC 1990); OSR-SW-152-00003, Operating Limits for the 340 Waste Handling Facility (WHC 1989); SD-RE-SAP-013, Safety Analysis Report for Packaging, Railroad Liquid Waste Tank Cars (Mercado 1993); SD-WM-TM-001, Safety Assessment Document for the 340 Waste Handling Facility (Berneski 1994a); SD-WM-SEL-016, 340 Facility Safety Equipment List (Berneski 1992); and 340 Complex Fire Hazard Analysis, Draft (Hughes Assoc. Inc. 1994)

  9. Liquefied Gaseous Fuels Spill Test Facility

    International Nuclear Information System (INIS)

    1993-02-01

    The US Department of Energy's liquefied Gaseous Fuels Spill Test Facility is a research and demonstration facility available on a user-fee basis to private and public sector test and training sponsors concerned with safety aspects of hazardous chemicals. Though initially designed to accommodate large liquefied natural gas releases, the Spill Test Facility (STF) can also accommodate hazardous materials training and safety-related testing of most chemicals in commercial use. The STF is located at DOE's Nevada Test Site near Mercury, Nevada, USA. Utilization of the Spill Test Facility provides a unique opportunity for industry and other users to conduct hazardous materials testing and training. The Spill Test Facility is the only facility of its kind for either large- or small-scale testing of hazardous and toxic fluids including wind tunnel testing under controlled conditions. It is ideally suited for test sponsors to develop verified data on prevention, mitigation, clean-up, and environmental effects of toxic and hazardous gaseous liquids. The facility site also supports structured training for hazardous spills, mitigation, and clean-up. Since 1986, the Spill Test Facility has been utilized for releases to evaluate the patterns of dispersion, mitigation techniques, and combustion characteristics of select materials. Use of the facility can also aid users in developing emergency planning under US P.L 99-499, the Superfund Amendments and Reauthorization Act of 1986 (SARA) and other regulations. The Spill Test Facility Program is managed by the US Department of Energy (DOE), Office of Fossil Energy (FE) with the support and assistance of other divisions of US DOE and the US Government. DOE/FE serves as facilitator and business manager for the Spill Test Facility and site. This brief document is designed to acquaint a potential user of the Spill Test Facility with an outline of the procedures and policies associated with the use of the facility

  10. 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

  11. AOV Facility Tool/Facility Safety Specifications -

    Data.gov (United States)

    Department of Transportation — Develop and maintain authorizing documents that are standards that facilities must follow. These standards are references of FAA regulations and are specific to the...

  12. Safety of magnetic fusion facilities: Requirements

    International Nuclear Information System (INIS)

    1996-05-01

    This Standard identifies safety requirements for magnetic fusion facilities. Safety functions are used to define outcomes that must be achieved to ensure that exposures to radiation, hazardous materials, or other hazards are maintained within acceptable limits. Requirements applicable to magnetic fusion facilities have been derived from Federal law, policy, and other documents. In addition to specific safety requirements, broad direction is given in the form of safety principles that are to be implemented and within which safety can be achieved

  13. Safety assessment methodologies for near surface disposal facilities. Results of a co-ordinated research project (ISAM). Volume 1: Review and enhancement of safety assessment approaches and tools. Volume 2: Test cases

    International Nuclear Information System (INIS)

    2004-07-01

    For several decades, countries have made use of near surface facilities for the disposal of low and intermediate level radioactive waste. In line with the internationally agreed principles of radioactive waste management, the safety of these facilities needs to be ensured during all stages of their lifetimes, including the post-closure period. By the mid 1990s, formal methodologies for evaluating the long term safety of such facilities had been developed, but intercomparison of these methodologies had revealed a number of discrepancies between them. Consequently, in 1997, the International Atomic Energy Agency launched a Co-ordinated Research Project (CRP) on Improvement of Safety Assessment Methodologies for Near Surface Disposal Facilities (ISAM). The particular objectives of the CRP were to provide a critical evaluation of the approaches and tools used in post-closure safety assessment for proposed and existing near-surface radioactive waste disposal facilities, enhance the approaches and tools used and build confidence in the approaches and tools used. The CRP ran until 2000 and resulted in the development of a harmonized assessment methodology (the ISAM project methodology), which was applied to a number of test cases. Over seventy participants from twenty-two Member States played an active role in the project and it attracted interest from around seven hundred persons involved with safety assessment in seventy-two Member States. The results of the CRP have contributed to the Action Plan on the Safety of Radioactive Waste Management which was approved by the Board of Governors and endorsed by the General Conference in September 2001. Specifically, they contribute to Action 5, which requests the IAEA Secretariat to 'develop a structured and systematic programme to ensure adequate application of the Agency's waste safety standards', by elaborating on the Safety Requirements on 'Near Surface Disposal of Radioactive Waste' (Safety Standards Series No. WS-R-1) and

  14. Universal Test Facility

    Science.gov (United States)

    Laughery, Mike

    A universal test facility (UTF) for Space Station Freedom is developed. In this context, universal means that the experimental rack design must be: automated, highly marketable, and able to perform diverse microgravity experiments according to NASA space station requirements. In order to fulfill these broad objectives, the facility's customers, and their respective requirements, are first defined. From these definitions, specific design goals and the scope of the first phase of this project are determined. An examination is first made into what types of research are most likely to make the UTF marketable. Based on our findings, the experiments for which the UTF would most likely be used included: protein crystal growth, hydroponics food growth, gas combustion, gallium arsenide crystal growth, microorganism development, and cell encapsulation. Therefore, the UTF is designed to fulfill all of the major requirements for the experiments listed above. The versatility of the design is achieved by taking advantage of the many overlapping requirements presented by these experiments.

  15. Application of RELAP5/MOD3.3 to Calculate Thermal Hydraulic Behavior of the Pressurizer Safety Valve Performance Test Facility

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-10-15

    The increase of the acceptance tolerance of Pressurizer Safety Valve (PSV) test is vital for the safe operation of nuclear power plants because the frequent tests may make the valves decrepit and become a cause of leak. Recently, Korea Hydro and Nuclear Power Company (KHNP) is building a PSV performance test facility to provide the technical background data for the relaxation of the acceptance tolerance of PSV including the valve pop-up characteristics and the loop seal dynamics (if the plant has the loop seal in the upstream of PSV). The discharge piping and supports must be designed to withstand severe transient hydrodynamic loads when the safety valve actuates. The evaluation of hydrodynamic loads is a two-step process: first the thermal hydraulic behavior in the piping must be defined, and then the hydrodynamic loads are calculated from the thermal hydraulic parameters such as pressure and mass flow. The hydrodynamic loads are used as input to the structural analysis.

  16. TESLA Test Facility. Status

    International Nuclear Information System (INIS)

    Aune, B.

    1996-01-01

    The TESLA Test Facility (TTF), under construction at DESY by an international collaboration, is an R and D test bed for the superconducting option for future linear e+/e-colliders. It consists of an infrastructure to process and test the cavities and of a 500 MeV linac. The infrastructure has been installed and is fully operational. It includes a complex of clean rooms, an ultra-clean water plant, a chemical etching installation and an ultra-high vacuum furnace. The linac will consist of four cryo-modules, each containing eight 1 meter long nine-cell cavities operated at 1.3 GHz. The base accelerating field is 15 MV/m. A first injector will deliver a low charge per bunch beam, with the full average current (8 mA in pulses of 800 μs). A more powerful injector based on RF gun technology will ultimately deliver a beam with high charge and low emittance to allow measurements necessary to qualify the TESLA option and to demonstrate the possibility of operating a free electron laser based on the Self-Amplified-Spontaneous-Emission principle. Overview and status of the facility will be given. Plans for the future use of the linac are presented. (R.P.)

  17. CLIC Test Facility 3

    CERN Multimedia

    Kossyvakis, I; Faus-golfe, A

    2007-01-01

    The design of CLIC is based on a two-beam scheme, where short pulses of high power 30 GHz RF are extracted from a drive beam running parallel to the main beam. The 3rd generation CLIC Test Facility (CTF3) will demonstrate the generation of the drive beam with the appropriate time structure, the extraction of 30 GHz RF power from this beam, as well as acceleration of a probe beam with 30 GHz RF cavities. The project makes maximum use of existing equipment and infrastructure of the LPI complex, which became available after the closure of LEP.

  18. 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...

  19. 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

  20. 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.

  1. 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.

  2. Facilities management and industrial safety

    International Nuclear Information System (INIS)

    2003-06-01

    This book lists occupation safety and health acts with purpose, definition and management system of safety and health, enforcement ordinance of occupation safety and health acts and enforcement regulations such as general rules, safety and health cover, system of management on safety and health, regulation of management on safety and health, regulations of harmfulness and protection of danger, heath management for workers, supervisor and command and inspection of machine and equipment.

  3. 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.

  4. 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

  5. Eccentric Coil Test Facility (ECTF)

    International Nuclear Information System (INIS)

    Burn, P.B.; Walstrom, P.L.; Anderson, W.C.; Marguerat, E.F.

    1975-01-01

    The conceptual design of a facility for testing superconducting coils under some conditions peculiar to tokamak systems is given. A primary element of the proposed facility is a large 25 MJ background solenoid. Discussions of the mechanical structure, the stress distribution and the thermal stability for this coil are included. The systems for controlling the facility and diagnosing test coil behavior are also described

  6. LMFBR safety experiment facility planning and analysis

    International Nuclear Information System (INIS)

    Stevenson, M.G.; Scott, J.H.

    1976-01-01

    In the past two years considerable effort has been placed on the planning and design of new facilities for the resolution of LMFBR safety issues. The paper reviews the key issues, the experiments needed to resolve them, and the design aspects of proposed new facilities. In addition, it presents a decision theory approach to selecting an optimal combination of modified and new facilities

  7. Arc Heated Scramjet Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Arc Heated Scramjet Test Facility is an arc heated facility which simulates the true enthalpy of flight over the Mach number range of about 4.7 to 8 for free-jet...

  8. 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

  9. Radioactive wastes. Safety of storage facilities

    International Nuclear Information System (INIS)

    Devillers, Ch.

    2001-01-01

    A radioactive waste storage facility is designed in a way that ensures the isolation of wastes with respect to the biosphere. This function comprises the damping of the gamma and neutron radiations from the wastes, and the confinement of the radionuclides content of the wastes. The safety approach is based on two time scales: the safety of the insulation system during the main phase of radioactive decay, and the assessment of the radiological risks following this phase. The safety of a surface storage facility is based on a three-barrier concept (container, storage structures, site). The confidence in the safety of the facility is based on the quality assurance of the barriers and on their surveillance and maintenance. The safety of a deep repository will be based on the site quality, on the design and construction of structures and on the quality of the safety demonstration. This article deals with the safety approach and principles of storage facilities: 1 - recall of the different types of storage facilities; 2 - different phases of the life of a storage facility and regulatory steps; 3 - safety and radiation protection goals (time scales, radiation protection goals); 4 - safety approach and principles of storage facilities: safety of the isolation system (confinement system, safety analysis, scenarios, radiological consequences, safety principles), assessment of the radiation risks after the main phase of decay; 5 - safety of surface storage facilities: safety analysis of the confinement system of the Aube plant (barriers, scenarios, modeling, efficiency), evaluation of radiological risks after the main phase of decay; experience feedback of the Manche plant; variants of surface storage facilities in France and abroad (very low activity wastes, mine wastes, short living wastes with low and average activity); 6 - safety of deep geological disposal facilities: legal framework of the French research; international context; safety analysis of the confinement system

  10. Engineering test facility design definition

    Science.gov (United States)

    Bercaw, R. W.; Seikel, G. R.

    1980-01-01

    The Engineering Test Facility (ETF) is the major focus of the Department of Energy (DOE) Magnetohydrodynamics (MHD) Program to facilitate commercialization and to demonstrate the commercial operability of MHD/steam electric power. The ETF will be a fully integrated commercial prototype MHD power plant with a nominal output of 200 MW sub e. Performance of this plant is expected to meet or surpass existing utility standards for fuel, maintenance, and operating costs; plant availability; load following; safety; and durability. It is expected to meet all applicable environmental regulations. The current design concept conforming to the general definition, the basis for its selection, and the process which will be followed in further defining and updating the conceptual design.

  11. Hot Cell Facility (HCF) Safety Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    MITCHELL,GERRY W.; LONGLEY,SUSAN W.; PHILBIN,JEFFREY S.; MAHN,JEFFREY A.; BERRY,DONALD T.; SCHWERS,NORMAN F.; VANDERBEEK,THOMAS E.; NAEGELI,ROBERT E.

    2000-11-01

    This Safety Analysis Report (SAR) is prepared in compliance with the requirements of DOE Order 5480.23, Nuclear Safety Analysis Reports, and has been written to the format and content guide of DOE-STD-3009-94 Preparation Guide for U. S. Department of Energy Nonreactor Nuclear Safety Analysis Reports. The Hot Cell Facility is a Hazard Category 2 nonreactor nuclear facility, and is operated by Sandia National Laboratories for the Department of Energy. This SAR provides a description of the HCF and its operations, an assessment of the hazards and potential accidents which may occur in the facility. The potential consequences and likelihood of these accidents are analyzed and described. Using the process and criteria described in DOE-STD-3009-94, safety-related structures, systems and components are identified, and the important safety functions of each SSC are described. Additionally, information which describes the safety management programs at SNL are described in ancillary chapters of the SAR.

  12. Hot Cell Facility (HCF) Safety Analysis Report

    International Nuclear Information System (INIS)

    MITCHELL, GERRY W.; LONGLEY, SUSAN W.; PHILBIN, JEFFREY S.; MAHN, JEFFREY A.; BERRY, DONALD T.; SCHWERS, NORMAN F.; VANDERBEEK, THOMAS E.; NAEGELI, ROBERT E.

    2000-01-01

    This Safety Analysis Report (SAR) is prepared in compliance with the requirements of DOE Order 5480.23, Nuclear Safety Analysis Reports, and has been written to the format and content guide of DOE-STD-3009-94 Preparation Guide for U. S. Department of Energy Nonreactor Nuclear Safety Analysis Reports. The Hot Cell Facility is a Hazard Category 2 nonreactor nuclear facility, and is operated by Sandia National Laboratories for the Department of Energy. This SAR provides a description of the HCF and its operations, an assessment of the hazards and potential accidents which may occur in the facility. The potential consequences and likelihood of these accidents are analyzed and described. Using the process and criteria described in DOE-STD-3009-94, safety-related structures, systems and components are identified, and the important safety functions of each SSC are described. Additionally, information which describes the safety management programs at SNL are described in ancillary chapters of the SAR

  13. Safety assessment for spent fuel storage facilities

    International Nuclear Information System (INIS)

    1994-01-01

    This Safety Practice has been prepared as part of the IAEA's programme on the safety assessment of interim spent fuel storage facilities which are not an integral part of an operating nuclear power plant. This report provides general guidance on the safety assessment process, discussing both deterministic and probabilistic assessment methods. It describes the safety assessment process for normal operation and anticipated operational occurrences and also related to accident conditions. 10 refs, 2 tabs

  14. Buffet test in the National Transonic Facility

    Science.gov (United States)

    Young, Clarence P., Jr.; Hergert, Dennis W.; Butler, Thomas W.; Herring, Fred M.

    1992-01-01

    A buffet test of a commercial transport model was accomplished in the National Transonic Facility at the NASA Langley Research Center. This aeroelastic test was unprecedented for this wind tunnel and posed a high risk to the facility. This paper presents the test results from a structural dynamics and aeroelastic response point of view and describes the activities required for the safety analysis and risk assessment. The test was conducted in the same manner as a flutter test and employed onboard dynamic instrumentation, real time dynamic data monitoring, automatic, and manual tunnel interlock systems for protecting the model. The procedures and test techniques employed for this test are expected to serve as the basis for future aeroelastic testing in the National Transonic Facility. This test program was a cooperative effort between the Boeing Commercial Airplane Company and the NASA Langley Research Center.

  15. Radiation safety training for accelerator facilities

    International Nuclear Information System (INIS)

    Trinoskey, P.A.

    1997-02-01

    In November 1992, a working group was formed within the U.S. Department of Energy's (DOE's) accelerator facilities to develop a generic safety training program to meet the basic requirements for individuals working in accelerator facilities. This training, by necessity, includes sections for inserting facility-specific information. The resulting course materials were issued by DOE as a handbook under its technical standards in 1996. Because experimenters may be at a facility for only a short time and often at odd times during the day, the working group felt that computer-based training would be useful. To that end, Lawrence Livermore National Laboratory (LLNL) and Argonne National Laboratory (ANL) together have developed a computer-based safety training program for accelerator facilities. This interactive course not only enables trainees to receive facility- specific information, but time the training to their schedule and tailor it to their level of expertise

  16. Electromagnetic Interface Testing Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Electromagnetic Interface Testing facilitysupports such testing asEmissions, Field Strength, Mode Stirring, EMP Pulser, 4 Probe Monitoring/Leveling System, and...

  17. Occupational Safety Review of High Technology Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Lee Cadwallader

    2005-01-31

    This report contains reviews of operating experiences, selected accident events, and industrial safety performance indicators that document the performance of the major US DOE magnetic fusion experiments and particle accelerators. These data are useful to form a basis for the occupational safety level at matured research facilities with known sets of safety rules and regulations. Some of the issues discussed are radiation safety, electromagnetic energy exposure events, and some of the more widespread issues of working at height, equipment fires, confined space work, electrical work, and other industrial hazards. Nuclear power plant industrial safety data are also included for comparison.

  18. Safety of Nuclear Fuel Cycle Facilities. Safety Requirements (Arabic Edition)

    International Nuclear Information System (INIS)

    2015-01-01

    This publication covers the broad scope of requirements for fuel cycle facilities that, in light of the experience and present state of technology, must be satisfied to ensure safety for the lifetime of the facility. Topics of specific relevance include aspects of nuclear fuel generation, storage, reprocessing and disposal

  19. Implementing partnerships in nonreactor facility safety analyses

    International Nuclear Information System (INIS)

    Courtney, J.C.; Perry, W.H.; Phipps, R.D.

    1996-01-01

    Faculty and students from LSU have been participating in nuclear safety analyses and radiation protection projects at ANL-W at INEL since 1973. A mutually beneficial relationship has evolved that has resulted in generation of safety-related studies acceptable to Argonne and DOE, NRC, and state regulatory groups. Most of the safety projects have involved the Hot Fuel Examination Facility or the Fuel Conditioning Facility; both are hot cells that receive spent fuel from EBR-II. A table shows some of the major projects at ANL-W that involved LSU students and faculty

  20. EMI Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — FUNCTION: Supports electromagnetic interference/radio frequency interference (EMI/RFI) testing of flight hardware. It is also used to support custom RF testing up to...

  1. Static Loads Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — FUNCTION: Provides the capability to perform large-scale structural loads testing on spacecraft and other structures. Results from these tests can be used to verify...

  2. Upgrading safety systems of industrial irradiation facilities

    International Nuclear Information System (INIS)

    Gomes, R.S.; Gomes, J.D.R.L.; Costa, E.L.C.; Costa, M.L.L.; Thomé, Z.D.

    2017-01-01

    The first industrial irradiation facility in operation in Brazil was designed in the 70s. Nowadays, twelve commercial and research facilities are in operation and two already decommissioned. Minor modifications and upgrades, as sensors replacement, have been introduced in these facilities, in order to reduce the technological gap in the control and safety systems. The safety systems are designed in agreement with the codes and standards at the time. Since then, new standards, codes and recommendations, as well as lessons learned from accidents, have been issued by various international committees or regulatory bodies. The rapid advance of the industry makes the safety equipment used in the original construction become obsolete. The decreasing demand for these older products means that they are no longer produced, which can make it impossible or costly to obtain spare parts and the expansion of legacy systems to include new features. This work aims to evaluate existing safety systems at Brazilian irradiation facilities, mainly the oldest facilities, taking into account the recommended IAEA's design requirements. Irrespective of the fact that during its operational period no event with victims have been recorded in Brazilian facilities, and that the regulatory inspections do not present any serious deviations regarding the safety procedures, it is necessary an assessment of safety system with the purpose of bringing their systems to 'the state of the art', avoiding their rapid obsolescence. This study has also taken into account the knowledge, concepts and solutions developed to upgrading safety system in irradiation facilities throughout the world. (author)

  3. Upgrading safety systems of industrial irradiation facilities

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, R.S.; Gomes, J.D.R.L.; Costa, E.L.C.; Costa, M.L.L., E-mail: rogeriog@cnen.gov.br, E-mail: jlopes@cnen.gov.br, E-mail: evaldo@cnen.gov.br, E-mail: mara@cnen.gov.br [Comissão Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil). Diretoria de Radioproteção e Segurança Nuclear; Thomé, Z.D., E-mail: zielithome@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Seção de Engenharia Nuclear

    2017-07-01

    The first industrial irradiation facility in operation in Brazil was designed in the 70s. Nowadays, twelve commercial and research facilities are in operation and two already decommissioned. Minor modifications and upgrades, as sensors replacement, have been introduced in these facilities, in order to reduce the technological gap in the control and safety systems. The safety systems are designed in agreement with the codes and standards at the time. Since then, new standards, codes and recommendations, as well as lessons learned from accidents, have been issued by various international committees or regulatory bodies. The rapid advance of the industry makes the safety equipment used in the original construction become obsolete. The decreasing demand for these older products means that they are no longer produced, which can make it impossible or costly to obtain spare parts and the expansion of legacy systems to include new features. This work aims to evaluate existing safety systems at Brazilian irradiation facilities, mainly the oldest facilities, taking into account the recommended IAEA's design requirements. Irrespective of the fact that during its operational period no event with victims have been recorded in Brazilian facilities, and that the regulatory inspections do not present any serious deviations regarding the safety procedures, it is necessary an assessment of safety system with the purpose of bringing their systems to 'the state of the art', avoiding their rapid obsolescence. This study has also taken into account the knowledge, concepts and solutions developed to upgrading safety system in irradiation facilities throughout the world. (author)

  4. 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

  5. Enhancement of safety for reprocessing facilities

    International Nuclear Information System (INIS)

    2012-06-01

    The adequacy of the safety measures for utility loss accidents in nuclear fuel reprocessing facilities which have been formulated by the nuclear enterprises is investigated in JNES which organizes an advanced committee to specifically study this problem. The results are reviewed in the present report including the case of such severe accidents as in Fukushima Daiichi Nuclear Power Plant. The report also represents a tentative proposal for examination standards of such unimaginable severe accidents as 'station blackout,' urgent safety measures necessary for reoperation of nuclear power plants and requested by nuclear and industrial safety agency, and pointing out and clarification of the potential weakness from the safety point of view, and collective and composite evaluation of safety of the relevant facilities. Furthermore, the definition of accident management is given as of controlled condition and the authorized way of thinking for the cases of plural events happening at the same time and the cases when risks exist radioactivity emits with explosion. (S. Ohno)

  6. Chemical process safety at fuel cycle facilities

    International Nuclear Information System (INIS)

    Ayres, D.A.

    1997-08-01

    This NUREG provides broad guidance on chemical safety issues relevant to fuel cycle facilities. It describes an approach acceptable to the NRC staff, with examples that are not exhaustive, for addressing chemical process safety in the safe storage, handling, and processing of licensed nuclear material. It expounds to license holders and applicants a general philosophy of the role of chemical process safety with respect to NRC-licensed materials; sets forth the basic information needed to properly evaluate chemical process safety; and describes plausible methods of identifying and evaluating chemical hazards and assessing the adequacy of the chemical safety of the proposed equipment and facilities. Examples of equipment and methods commonly used to prevent and/or mitigate the consequences of chemical incidents are discussed in this document

  7. Solenoid Testing Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Current Configuration: Accommodate a device under test up to 2.8 m diameter, 0.7 m height and 15,000 lbs. weight. Up to 10 g/s, 4.5 K helium flow. Up to 250 A test...

  8. Airborne Test Bed Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Laboratory operates the main hangar on the Hanscom Air Force Base flight line. This very large building (~93,000sqft) accommodates the Laboratory's airborne test...

  9. 33-GVA interrupter test facility

    International Nuclear Information System (INIS)

    Parsons, W.M.; Honig, E.M.; Warren, R.W.

    1979-01-01

    The use of commercial ac circuit breakers for dc switching operations requires that they be evaluated to determine their dc limitations. Two 2.4-GVA facilities have been constructed and used for this purpose at LASL during the last several years. In response to the increased demand on switching technology, a 33-GVA facility has been constructed. Novel features incorporated into this facility include (1) separate capacitive and cryogenic inductive energy storage systems, (2) fiber-optic controls and optically-coupled data links, and (3) digital data acquisition systems. Facility details and planned tests on an experimental rod-array vacuum interrupter are presented

  10. FAST FLUX TEST FACILITY DRIVER FUEL MEETING

    Energy Technology Data Exchange (ETDEWEB)

    None,

    1966-06-01

    The Pacific Northwest Laboratory has convened this meeting to enlist the best talents of our laboratories and industry in soliciting factual, technical information pertinent to the Pacific Northwest's Laboratory's evaluation of the potential fuel systems for the Fast Flux Test Facility. The particular factors emphasized for these fuel systems are those associated with safety, ability to meet testing objectives, and economics. The proceedings includes twenty-three presentations, along with a transcript of the discussion following each, as well as a summary discussion.

  11. Reverberant Acoustic Test Facility (RATF)

    Data.gov (United States)

    Federal Laboratory Consortium — The very large Reverberant Acoustic Test Facility (RATF) at the NASA Glenn Research Center (GRC), Plum Brook Station, is currently under construction and is due to...

  12. Elevated Fixed Platform Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Elevated Fixed Platform (EFP) is a helicopter recovery test facility located at Lakehurst, NJ. It consists of a 60 by 85 foot steel and concrete deck built atop...

  13. LLNL superconducting magnets test facility

    Energy Technology Data Exchange (ETDEWEB)

    Manahan, R; Martovetsky, N; Moller, J; Zbasnik, J

    1999-09-16

    The FENIX facility at Lawrence Livermore National Laboratory was upgraded and refurbished in 1996-1998 for testing CICC superconducting magnets. The FENIX facility was used for superconducting high current, short sample tests for fusion programs in the late 1980s--early 1990s. The new facility includes a 4-m diameter vacuum vessel, two refrigerators, a 40 kA, 42 V computer controlled power supply, a new switchyard with a dump resistor, a new helium distribution valve box, several sets of power leads, data acquisition system and other auxiliary systems, which provide a lot of flexibility in testing of a wide variety of superconducting magnets in a wide range of parameters. The detailed parameters and capabilities of this test facility and its systems are described in the paper.

  14. Freshwater Treatment and Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Freshwater Treatment and Test Facility, located at SANGB, has direct year-round access to water from Lake St. Clair and has a State of Michigan approved National...

  15. Mirror fusion test facility

    International Nuclear Information System (INIS)

    Post, R.F.

    1978-01-01

    The MFTF is a large new mirror facility under construction at Livermore for completion in 1981--82. It represents a scaleup, by a factor of 50 in plasma volume, a factor of 5 or more in ion energy, and a factor of 4 in magnetic field intensity over the Livermore 2XIIB experiment. Its magnet, employing superconducting NbTi windings, is of Yin-Yang form and will weigh 200 tons. MFTF will be driven by neutral beams of two levels of current and energy: 1000 amperes of 20 keV (accelerating potential) pulsed beams for plasma startup; 750 amperes of 80 keV beams of 0.5 second duration for temperature buildup and plasma sustainment. Two operating modes for MFTF are envisaged: The first is operation as a conventional mirror cell with n/sup tau/ approximately equal to 10 12 cm -3 sec, W/sub i/ = 50 keV, where the emphasis will be on studying the physics of mirror cells, particularly the issues of improved techniques of stabilization against ion cyclotron modes and of maximization of the electron temperature. The second possible mode is the further study of the Field Reversed Mirror idea, using high current neutral beams to sustain the field-reversed state. Anticipating success in the coming Livermore Tandem Mirror Experiment (TMX) MFTF has been oriented so that it could comprise one end cell of a scaled up TM experiment. Also, if MFTF were to succeed in achieving a FR state it could serve as an essentially full-sized physics prototype of one cell of a FRM fusion power plant

  16. Safety assessment for radioactive waste disposal facility

    International Nuclear Information System (INIS)

    Thanaletchumy Karuppiah; Mohd Abdul Wahab Yusof; Nik Marzuki Nik Ibrahim; Nurul Wahida Ahmad Khairuddin

    2008-08-01

    Safety assessments are used to evaluate the performance of a radioactive waste disposal facility and its impact on human health and the environment. This paper presents the overall information and methodology to carry out the safety assessment for a long term performance of a disposal system. A case study was also conducted to gain hands-on experience in the development and justification of scenarios, the formulation and implementation of models and the analysis of results. AMBER code using compartmental modeling approach was used to represent the migration and fate of contaminants in this training. This safety assessment is purely illustrative and it serves as a starting point for each development stage of a disposal facility. This assessment ultimately becomes more detail and specific as the facility evolves. (Author)

  17. Laser safety at high profile laser facilities

    International Nuclear Information System (INIS)

    Barat, K.

    2010-01-01

    Complete text of publication follows. Laser safety has been an active concern of laser users since the invention of the laser. Formal standards were developed in the early 1970's and still continue to be developed and refined. The goal of these standards is to give users guidance on the use of laser and consistent safety guidance and requirements for laser manufacturers. Laser safety in the typical research setting (government laboratory or university) is the greatest challenge to the laser user and laser safety officer. This is due to two factors. First, the very nature of research can put the user at risk; consider active manipulation of laser optics and beam paths, and user work with energized systems. Second, a laser safety culture that seems to accept laser injuries as part of the graduate student educational process. The fact is, laser safety at research settings, laboratories and universities still has long way to go. Major laser facilities have taken a more rigid and serious view of laser safety, its controls and procedures. Part of the rationale for this is that these facilities draw users from all around the world presenting the facility with a work force of users coming from a wide mix of laser safety cultures. Another factor is funding sources do not like bad publicity which can come from laser accidents and a poor safety record. The fact is that injuries, equipment damage and lost staff time slow down progress. Hence high profile/large laser projects need to adapt a higher safety regimen both from an engineering and administrative point of view. This presentation will discuss all these points and present examples. Acknowledgement. This work has been supported by the University of California, Director, Office of Science.

  18. Preliminary safety assessment of the WIPP facility

    International Nuclear Information System (INIS)

    Balestri, R.J.; Torres, B.W.; Pahwa, S.B.; Brannen, J.P.

    1979-01-01

    This paper summarizes the efforts to perform a safety assessment of the Waste Isolation Pilot Plant (WIPP) facility being proposed for southeastern New Mexico. This preliminary safety assessment is limited to a consequence assessment in terms of the dose to a maximally exposed individual as a result of introducing the radionuclides into the biosphere. The extremely low doses to the organs as a result of the liquid breach scenarios are contrasted with the background radiation

  19. Mechanistic facility safety and source term analysis

    International Nuclear Information System (INIS)

    PLYS, M.G.

    1999-01-01

    A PC-based computer program was created for facility safety and source term analysis at Hanford The program has been successfully applied to mechanistic prediction of source terms from chemical reactions in underground storage tanks, hydrogen combustion in double contained receiver tanks, and proccss evaluation including the potential for runaway reactions in spent nuclear fuel processing. Model features include user-defined facility room, flow path geometry, and heat conductors, user-defined non-ideal vapor and aerosol species, pressure- and density-driven gas flows, aerosol transport and deposition, and structure to accommodate facility-specific source terms. Example applications are presented here

  20. Safety culture in industrial radiography facility

    International Nuclear Information System (INIS)

    Vincent-Furo, Evelyn

    2015-02-01

    This project reviewed published IAEA materials and other documents on safety culture with specific references to industrial radiography. Safety culture requires all duties important to safety to be carried out correctly, with alertness, due thought and full knowledge, sound judgment and a proper sense of accountability. The development and maintenance of safety culture in an operating organization has to cover management systems, policies, responsibilities, procedures and organizational arrangements. The essence is to control radiation hazard, optimize radiation protection to prevent or reduce exposures and mitigate the consequences of accidents and incidents. To achieve a high degree of safety culture appropriate national and international infrastructure should exist to ensure effective training of workers and management system that supports commitment to safety culture at all level of the organization; management, managers and workforce. The result of the review revealed that all accidents in industrial radiography facilities were due to poor safety culture practices including inadequate regulatory control oversight. Some recommendations are provided and if implemented could improve safety culture leading to good safety performance which will significantly reduce accidents and their consequences in industrial radiography. These examples call for a review of safety culture in Industrial radiography. (au)

  1. Operational and safety requirement of radiation facility

    International Nuclear Information System (INIS)

    Zulkafli Ghazali

    2007-01-01

    Gamma and electron irradiation facilities are the most common industrial sources of ionizing radiation. They have been used for medical, industrial and research purposes since the 1950s. Currently there are more than 160 gamma irradiation facilities and over 600 electron beam facilities in operation worldwide. These facilities are either used for the sterilization of medical and pharmaceutical products, the preservation of foodstuffs, polymer synthesis and modification, or the eradication of insect infestation. Irradiation with electron beam, gamma ray or ultra violet light can also destroy complex organic contaminants in both liquid and gaseous waste. EB systems are replacing traditional chemical sterilization methods in the medical supply industry. The ultra-violet curing facility, however, has found more industrial application in printing and furniture industries. Gamma and electron beam facilities produce very high dose rates during irradiation, and thus there is a potential of accidental exposure in the irradiation chamber which can be lethal within minutes. Although, the safety record of this industry has been relatively very good, there have been fatalities recorded in Italy (1975), Norway (1982), El Salvador (1989) and Israel (1990). Precautions against uncontrolled entry into irradiation chamber must therefore be taken. This is especially so in the case of gamma irradiation facilities those contain large amounts of radioactivity. If the mechanism for retracting the source is damaged, the source may remain exposed. This paper will, to certain extent, describe safety procedure and system being installed at ALURTRON, Nuclear Malaysia to eliminate accidental exposure of electron beam irradiation. (author)

  2. 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

  3. 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.

  4. 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)

  5. 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

  6. (abstract) Cryogenic Telescope Test Facility

    Science.gov (United States)

    Luchik, T. S.; Chave, R. G.; Nash, A. E.

    1995-01-01

    An optical test Dewar is being constructed with the unique capability to test mirrors of diameter less than or equal to 1 m, f less than or equal to 6, at temperatures from 300 to 4.2 K with a ZYGO Mark IV interferometer. The design and performance of this facility will be presented.

  7. 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

  8. Corrosion testing facilities in India

    International Nuclear Information System (INIS)

    Viswanathan, R.; Subramanian, Venu

    1981-01-01

    Major types of corrosion tests, establishment of specifications on corrosion testing and scope of their application in practice are briefly described. Important organizations in the world which publish specifications/standards are listed. Indian organizations which undertake corrosion testing and test facilities available at them are also listed. Finally in an appendix, a comprehensive list of specifications relevant to corrosion testing is given. It is arranged under the headings: environmental testing, humidity tests, salt spray/fog tests, immersion tests, specification corrosion phenomena, (tests) with respect to special corrosion media, (tests) with respect to specific corrosion prevention methods, and specific corrosion tests using electrical and electrochemical methods (principles). Each entry in the list furnishes information about: nature of the test, standard number, and its specific application. (M.G.B.)

  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. 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)

  11. ACP Facility Safety Surveillance System Installation

    International Nuclear Information System (INIS)

    You, Gil Sung; Kook, D. H.; Choung, W. M.; Ku, J. H.; Cho, I. J.; You, G. S.; Kwon, K. C.; Lee, W. K.; Lee, E. P.

    2006-10-01

    The Advanced spent fuel Conditioning Process is under development for effective management of spent fuel by converting UO 2 into U-metal. For demonstration of this process, α-γ type new hotcell was built in the IMEF basement. All facilities which treat radioactive materials must manage CCTV system which is under control of Health Physics department. Three main points (including hotcell rear door area) have each camera, but operators who are in charge of facility management need to check the safety of the facility immediately through the network in his office. This needs introduce additional network cameras installation and this new surveillance system is expected to update the whole safety control ability with existing system

  12. National Ignition Facility Project Site Safety Program

    International Nuclear Information System (INIS)

    Dun, C

    2003-01-01

    This Safety Program for the National Ignition Facility (NIF) presents safety protocols and requirements that management and workers shall follow to assure a safe and healthful work environment during activities performed on the NIF Project site. The NIF Project Site Safety Program (NPSSP) requires that activities at the NIF Project site be performed in accordance with the ''LLNL ES and H Manual'' and the augmented set of controls and processes described in this NIF Project Site Safety Program. Specifically, this document: (1) Defines the fundamental NIF site safety philosophy. (2) Defines the areas covered by this safety program (see Appendix B). (3) Identifies management roles and responsibilities. (4) Defines core safety management processes. (5) Identifies NIF site-specific safety requirements. This NPSSP sets forth the responsibilities, requirements, rules, policies, and regulations for workers involved in work activities performed on the NIF Project site. Workers are required to implement measures to create a universal awareness that promotes safe practice at the work site and will achieve NIF management objectives in preventing accidents and illnesses. ES and H requirements are consistent with the ''LLNL ES and H Manual''. This NPSSP and implementing procedures (e.g., Management Walkabout, special work procedures, etc.,) are a comprehensive safety program that applies to NIF workers on the NIF Project site. The NIF Project site includes the B581/B681 site and support areas shown in Appendix B

  13. Radiation safety in X-ray facilities

    International Nuclear Information System (INIS)

    2001-09-01

    The guide specifies the radiation safety requirements for structural shielding and other safety arrangements used in X-ray facilities in medical and veterinary X-ray activities and in industry, research and education. The guide is also applicable to premises in which X-ray equipment intended for radiation therapy and operating at a voltage of less than 25 kV is used. The guide applies to new X-ray facilities in which X-ray equipment that has been used elsewhere is transferred. The radiation safety requirements for radiation therapy X-ray devices operating at a voltage exceeding 25 kV, and for the premices in which such devices are used, are set out in Guide ST 2.2

  14. Radiation safety in X-ray facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-09-01

    The guide specifies the radiation safety requirements for structural shielding and other safety arrangements used in X-ray facilities in medical and veterinary X-ray activities and in industry, research and education. The guide is also applicable to premises in which X-ray equipment intended for radiation therapy and operating at a voltage of less than 25 kV is used. The guide applies to new X-ray facilities in which X-ray equipment that has been used elsewhere is transferred. The radiation safety requirements for radiation therapy X-ray devices operating at a voltage exceeding 25 kV, and for the premices in which such devices are used, are set out in Guide ST 2.2.

  15. 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

  16. 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

  17. The Brookhaven Accelerator Test Facility

    International Nuclear Information System (INIS)

    Batchelor, K.; Ben-Zvi, I.; Fernow, R.C.; Fischer, J.; Fisher, A.S.; Gallardo, J.; Jialin, Xie; Kirk, H.G.; Parsa, Z.; Palmer, R.B.; Rao, T.; Rogers, J.; Sheehan, J.; Tsang, T.Y.F.; Ulc, S.; Van Steenbergen, A.; Woodle, M.; Zhang, R.S.; McDonald, K.T.; Russell, D.P.; Jiang, Z.Y.; Pellegrini, C.; Wang, X.J.

    1990-01-01

    The Accelerator Test Facility (ATF), presently under construction at Brookhaven National laboratory, is described. It consists of a 50-MeV electron beam synchronizable to a high-peak power CO 2 laser. The interaction of electrons with the laser field will be probed, with some emphasis on exploring laser-based acceleration techniques. 5 refs., 2 figs

  18. Criticality safety and facility design considerations

    International Nuclear Information System (INIS)

    Waltz, W.R.

    1991-06-01

    Operations with fissile material introduce the risk of a criticality accident that may be lethal to nearby personnel. In addition, concerns over criticality safety can result in substantial delays and shutdown of facility operations. For these reasons, it is clear that the prevention of a nuclear criticality accident should play a major role in the design of a nuclear facility. The emphasis of this report will be placed on engineering design considerations in the prevention of criticality. The discussion will not include other important aspects, such as the physics of calculating limits nor criticality alarm systems

  19. Fusion Materials Irradiation Test Facility

    International Nuclear Information System (INIS)

    Kemp, E.L.; Trego, A.L.

    1979-01-01

    A Fusion Materials Irradiation Test Facility is being designed to be constructed at Hanford, Washington, The system is designed to produce about 10 15 n/cm-s in a volume of approx. 10 cc and 10 14 n/cm-s in a volume of 500 cc. The lithium and target systems are being developed and designed by HEDL while the 35-MeV, 100-mA cw accelerator is being designed by LASL. The accelerator components will be fabricated by US industry. The total estimated cost of the FMIT is $105 million. The facility is scheduled to begin operation in September 1984

  20. Aircraft Test & Evaluation Facility (Hush House)

    Data.gov (United States)

    Federal Laboratory Consortium — The Aircraft Test and Evaluation Facility (ATEF), or Hush House, is a noise-abated ground test sub-facility. The facility's controlled environment provides 24-hour...

  1. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    Energy Technology Data Exchange (ETDEWEB)

    C.E. Sanders

    2005-04-07

    This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the calculation is to demonstrate that the handling operations of canisters performed in the CHF 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 ''Canister Handling Facility Description Document'' (BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dual-purpose canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for

  2. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    International Nuclear Information System (INIS)

    C.E. Sanders

    2005-01-01

    This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the calculation is to demonstrate that the handling operations of canisters performed in the CHF 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 ''Canister Handling Facility Description Document'' (BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dual-purpose canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for the current design of the CHF and may not reflect the ongoing design evolution of the facility

  3. Safety problems with abandoned explosive facilities

    International Nuclear Information System (INIS)

    Courtright, W.C.

    1969-01-01

    Procedures were developed for the safe removal of explosive and radioactive contaminated materials structures and drains from abandoned sites, including explosives processing and service buildings with a goal to return the entire area to its natural state and to permit public access. The safety problems encountered in the cleanup and their solutions are applicable to modification and maintenance work in operating explosive facilities. (U.S.)

  4. Cold Vacuum Drying (CVD) Facility Technical Safety Requirements

    International Nuclear Information System (INIS)

    KRAHN, D.E.

    2000-01-01

    The Technical Safety Requirements (TSRs) for the Cold Vacuum Drying Facility define acceptable conditions, safe boundaries, bases thereof, and management or administrative controls required to ensure safe operation during receipt of multi-canister overpacks (MCOs) containing spent nuclear fuel. removal of free water from the MCOs using the cold vacuum drying process, and inerting and testing of the MCOs before transport to the Canister Storage Building. Controls required for public safety, significant defense in depth, significant worker safety, and for maintaining radiological and toxicological consequences below risk evaluation guidelines are included

  5. Characterizing experiments of the PPOOLEX test facility

    Energy Technology Data Exchange (ETDEWEB)

    Puustinen, M.; Laine, J. (Lappeenranta Univ. of Technology, Nuclear Safety Research Unit (Finland))

    2008-07-15

    This report summarizes the results of the characterizing test series in 2007 with the scaled down PPOOLEX facility designed and constructed at Lappeenranta University of Technology. Air and steam/air mixture was blown into the dry well compartment and from there through a DN200 blowdown pipe to the condensation pool (wet well). Altogether eight air and four steam/air mixture experiments, each consisting of several blows (tests), were carried out. The main purpose of the experiment series was to study the general behavior of the facility and the performance of basic instrumentation. Proper operation of automation, control and safety systems was also tested. The test facility is a closed stainless steel vessel divided into two compartments, dry well and wet well. The facility is equipped with high frequency measurements for capturing different aspects of the investigated phenomena. The general behavior of the PPOOLEX facility differs significantly from that of the previous POOLEX facility because of the closed two-compartment structure of the test vessel. Heat-up by several tens of degrees due to compression in both compartments was the most obvious evidence of this. Temperatures also stratified. Condensation oscillations and chugging phenomenon were encountered in those tests where the fraction of non-condensables had time to decrease significantly. A radical change from smooth condensation behavior to oscillating one occurred quite abruptly when the air fraction of the blowdown pipe flow dropped close to zero. The experiments again demonstrated the strong diminishing effect that noncondensable gases have on dynamic unsteady loadings experienced by submerged pool structures. BWR containment like behavior related to the beginning of a postulated steam line break accident was observed in the PPOOLEX test facility during the steam/air mixture experiments. The most important task of the research project, to produce experimental data for code simulation purposes, can be

  6. Facility effluent monitoring plan for the fast flux test facility

    International Nuclear Information System (INIS)

    Nickels, J.M.; Dahl, N.R.

    1992-11-01

    A facility effluent monitoring plan is required by the US Department of Energy in US Department of Energy Order 5400.1 for any operations that involve hazardous materials and radioactive substances that could affect employee or public safety or the environment. A Facility Effluent Monitoring Plan determination was performed during calendar year 1991 and the evaluation requires the need for a facility effluent monitoring plan. This facility effluent monitoring plan assesses effluent monitoring systems and evaluates whether they are adequate to ensure the public health and safety as specified in applicable federal, state, and local requirements

  7. 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.

  8. Integration of radiation and physical safety in large radiator facilities

    International Nuclear Information System (INIS)

    Lima, P.P.M.; Benedito, A.M.; Lima, C.M.A.; Silva, F.C.A. da

    2017-01-01

    Growing international concern about radioactive sources after the Sept. 11, 2001 event has led to a strengthening of physical safety. There is evidence that the illicit use of radioactive sources is a real possibility and may result in harmful radiological consequences for the population and the environment. In Brazil there are about 2000 medical, industrial and research facilities with radioactive sources, of which 400 are Category 1 and 2 classified by the - International Atomic Energy Agency - AIEA, where large irradiators occupy a prominent position due to the very high cobalt-60 activities. The radiological safety is well established in these facilities, due to the intense work of the authorities in the Country. In the paper the main aspects on radiological and physical safety applied in the large radiators are presented, in order to integrate both concepts for the benefit of the safety as a whole. The research showed that the items related to radiation safety are well defined, for example, the tests on the access control devices to the irradiation room. On the other hand, items related to physical security, such as effective control of access to the company, use of safety cameras throughout the company, are not yet fully incorporated. Integration of radiation and physical safety is fundamental for total safety. The elaboration of a Brazilian regulation on the subject is of extreme importance

  9. Safety analysis and risk assessment of the National Ignition Facility

    International Nuclear Information System (INIS)

    Brereton, S.; McLouth, L.; Odell, B.

    1996-01-01

    The National Ignition Facility (NIF) is a proposed U.S. Department of Energy inertial confinement laser fusion facility. The candidate sites for locating the NIF are: Los Alamos National Laboratory, Sandia National Laboratory, the Nevada Test Site, and Lawrence Livermore National Laboratory (LLNL), the preferred site. The NIF will operate by focusing 192 laser beams onto a tiny deuterium-tritium target located at the center of a spherical target chamber. The NIF mission is to achieve inertial confinement fusion (ICF) ignition, access physical conditions in matter of interest to nuclear weapons physics, provide an above ground simulation capability for nuclear weapons effects testing, and contribute to the development of inertial fusion for electrical power production. The NIF has been classified as a radiological, low hazard facility on the basis of a preliminary hazards analysis and according to the DOE methodology for facility classification. This requires that a safety analysis be prepared under DOE Order 5481.1B, Safety Analysis and Review System. A draft Preliminary Safety Analysis Report (PSAR) has been written, and this will be finalized later in 1996. This paper summarizes the safety issues associated with the operation of the NIF and the methodology used to study them. It provides a summary of the methodology, an overview of the hazards, estimates maximum routine and accidental exposures for the preferred site of LLNL, and concludes that the risks from NIF operations are low

  10. Millimeter-wave Instrumentation Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Millimeter-wave Instrumentation Test Facility conducts basic research in propagation phenomena, remote sensing, and target signatures. The facility has a breadth...

  11. ITER primary cryopump test facility

    International Nuclear Information System (INIS)

    Petersohn, N.; Mack, A.; Boissin, J.C.; Murdoc, D.

    1998-01-01

    A cryopump as ITER primary vacuum pump is being developed at FZK under the European fusion technology programme. The ITER vacuum system comprises of 16 cryopumps operating in a cyclic mode which fulfills the vacuum requirements in all ITER operation modes. Prior to the construction of a prototype cryopump, the concept is tested on a reduced scale model pump. To test the model pump, the TIMO facility is being built at FZK in which the model pump operation under ITER environmental conditions, except for tritium exposure, neutron irradiation and magnetic fields, can be simulated. The TIMO facility mainly consists of a test vessel for ITER divertor duct simulation, a 600 W refrigerator system supplying helium in the 5 K stage and a 30 kW helium supply system for the 80 K stage. The model pump test programme will be performed with regard to the pumping performance and cryogenic operation of the pump. The results of the model pump testing will lead to the design of the full scale ITER cryopump. (orig.)

  12. The Safety and Tritium Applied Research (STAR) Facility: Status-2004

    International Nuclear Information System (INIS)

    Anderl, R.A.; Longhurst, G.R.; Pawelko, R.J.; Sharpe, J.P.; Schuetz, S.T.; Petti, D.A.

    2005-01-01

    The Safety and Tritium Applied Research (STAR) Facility, a US DOE National User Facility at the Idaho National Engineering and Environmental Laboratory (INEEL), comprises capabilities and infrastructure to support both tritium and non-tritium research activities important to the development of safe and environmentally friendly fusion energy. Research thrusts include (1) interactions of tritium and deuterium with plasma-facing-component (PFC) materials, (2) fusion safety issues [PFC material chemical reactivity and dust/debris generation, activation product mobilization, tritium behavior in fusion systems], and (3) molten salts and fusion liquids for tritium breeder and coolant applications. This paper updates the status of STAR and the capabilities for ongoing research activities, with an emphasis on the development, testing and integration of the infrastructure to support tritium research activities. Key elements of this infrastructure include a tritium storage and assay system, a tritium cleanup system to process glovebox and experiment tritiated effluent gases, and facility tritium monitoring systems

  13. Safety guide data on radiation shielding in a reprocessing facility

    International Nuclear Information System (INIS)

    Sekiguchi, Noboru; Naito, Yoshitaka

    1986-04-01

    In a reprocessing facility, various radiation sources are handled and have many geometrical conditions. To aim drawing up a safety guidebook on radiation shielding in order to evaluate shielding safety in a reprocessing facility with high reliability and reasonableness, JAERI trusted investigation on safety evaluation techniques of radiation shielding in a reprocessing facility to Nuclear Safety Research Association. This report is the collection of investigation results, and describes concept of shielding safety design principle, radiation sources in reprocessing facility and estimation of its strength, techniques of shielding calculations, and definite examples of shielding calculation in reprocessing facility. (author)

  14. 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

  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. Compressed Gas Safety for Experimental Fusion Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Lee C. Cadwallader

    2004-09-01

    Experimental fusion facilities present a variety of hazards to the operators and staff. There are unique or specialized hazards, including magnetic fields, cryogens, radio frequency emissions, and vacuum reservoirs. There are also more general industrial hazards, such as a wide variety of electrical power, pressurized air, and cooling water systems in use, there are crane and hoist loads, working at height, and handling compressed gas cylinders. This paper outlines the projectile hazard assoicated with compressed gas cylinders and mthods of treatment to provide for compressed gas safety. This information should be of interest to personnel at both magnetic and inertial fusion experiments.

  17. Safety tests file

    International Nuclear Information System (INIS)

    2011-01-01

    The design and operation of nuclear power plants is governed by strict and clearly defined regulations designed to ensure their safety in all circumstances. Since the first nuclear reactors were commissioned, the basic safety principles and the corresponding practical requirements have constantly evolved and been enhanced, benefiting from operating experience feedback from reactors around the world (about 500 production reactors currently in service). Reactor safety has from the outset been built around the 'defense in depth' concept, which aims to prevent melting of the core and radioactive releases into the environment. It can be summarized as follows: over and above all the measures taken to prevent accidents, the principle that accidents do occur has to be accepted. We then assess their consequences and take steps to contain them at the level of severity at which they occur. (authors)

  18. TFTR neutral-beam test facility

    International Nuclear Information System (INIS)

    Turitzin, N.M.; Newman, R.A.

    1981-11-01

    TFTR Neutral Beam System will have thirteen discharge ion sources, each with its own power supply. Twelve of these will be utilized for supplemental heating of the TFTR tokamak plasma, while the thirteenth will be dedicated to an off-machine test chamber for source development and/or conditioning. A test installation for one source was set up using prototype equipment to discover and correct possible deficiencies, and to properly coordinate the equipment. This test facility represents the first opportunity for assembling an integrated system of hardware supplied by diverse vendors, each of whom designed and built his equipment to performance specifications. For the installation and coordination of the different portions of the total system, particular attention was given to personnel safety and safe equipment operation. This paper discusses various system components, their characteristics, interconnection and control. Results of the recently initiated test phase will be reported at a later date

  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. Operation of the hot test loop facilities

    International Nuclear Information System (INIS)

    Cheong, Moon Ki; Park, Choon Kyeong; Won, Soon Yeon; Yang, Sun Kyu; Cheong, Jang Whan; Cheon, Se Young; Song, Chul Hwa; Jeon, Hyeong Kil; Chang, Suk Kyu; Jeong, Heung Jun; Cho, Young Ro; Kim, Bok Duk; Min, Kyeong Ho

    1994-12-01

    The objective of this project is to obtain the available experimental data and to develop the measuring techniques through taking full advantage of the facilities. The facilities operated by the thermal hydraulics department have been maintained and repaired in order to carry out the thermal hydraulics tests necessary for providing the available data. The performance tests for double grid type bottom end piece which was improved on the debris filtering effectivity were performed using the PWR-Hot Test Loop. The CANDU-Hot Test Loop was operated to carry out the pressure drop tests and strength tests of fuel. The Cold Test Loop was used to obtain the local velocity data in subchannel within fuel bundle and to understand the characteristic of pressure drop required for improving the nuclear fuel and to develop the advanced measuring techniques. RCS Loop, which is used to measure the CHF, is presently under design and construction. B and C Loop is designed and constructed to assess the automatic depressurization safety system behavior. 4 tabs., 79 figs., 7 refs. (Author) .new

  1. 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)

  2. Complementary safety assessment assessment of nuclear facilities - Tricastin facility - AREVA

    International Nuclear Information System (INIS)

    2011-01-01

    This complementary safety assessment analyses the robustness of the Areva part of the Tricastin nuclear site to extreme situations such as those that led to the Fukushima accident. This study includes the following facilities: Areva NC Pierrelatte, EURODIF production, Comurhex Pierrelatte, Georges Besse II plant and Socatri. Robustness is the ability for the plant to withstand events beyond which the plant was designed. Robustness is linked to safety margins but also to the situations leading to a sudden deterioration of the accidental sequence. Moreover, safety is not only a matter of design or engineered systems but also a matter of organizing: task organization (including subcontracting) as well as the setting of emergency plans or the inventory of nuclear materials are taken into consideration in this assessment. This report is divided into 10 main chapters: 1) the feedback experience of the Fukushima accident; 2) description of the site and its surroundings; 3) featuring of the site's activities and installations; 4) accidental sequences; 5) protection from earthquakes; 6) protection from floods; 7) protection from other extreme natural disasters; 8) the loss of electrical power and of the heat sink; 9) the management of severe accidents; and 10) subcontracting policy. This analysis has identified 5 main measures to be taken to limit the risks linked to natural disasters: -) continuing the program for replacing the current conversion plant and the enrichment plant; -) renewing the storage of hydrofluoric acid at the de-fluorination workshop; -) assessing the seismic behaviour of some parts of the de-fluorination workshop and of the fluorine fabrication workshop; -) improving the availability of warning and information means in case of emergency; and -) improving the means to mitigate accidental gaseous releases. (A.C.)

  3. Spallation Neutron Source Accelerator Facility Target Safety and Non-safety Control Systems

    International Nuclear Information System (INIS)

    Battle, Ronald E.; DeVan, B.; Munro, John K. Jr.

    2006-01-01

    The Spallation Neutron Source (SNS) is a proton accelerator facility that generates neutrons for scientific researchers by spallation of neutrons from a mercury target. The SNS became operational on April 28, 2006, with first beam on target at approximately 200 W. The SNS accelerator, target, and conventional facilities controls are integrated by standardized hardware and software throughout the facility and were designed and fabricated to SNS conventions to ensure compatibility of systems with Experimental Physics Integrated Control System (EPICS). ControlLogix Programmable Logic Controllers (PLCs) interface to instruments and actuators, and EPICS performs the high-level integration of the PLCs such that all operator control can be accomplished from the Central Control room using EPICS graphical screens that pass process variables to and from the PLCs. Three active safety systems were designed to industry standards ISA S84.01 and IEEE 603 to meet the desired reliability for these safety systems. The safety systems protect facility workers and the environment from mercury vapor, mercury radiation, and proton beam radiation. The facility operators operated many of the systems prior to beam on target and developed the operating procedures. The safety and non-safety control systems were tested extensively prior to beam on target. This testing was crucial to identify wiring and software errors and failed components, the result of which was few problems during operation with beam on target. The SNS has continued beam on target since April to increase beam power, check out the scientific instruments, and continue testing the operation of facility subsystems

  4. Assessment of public safety around EDF facilities

    Energy Technology Data Exchange (ETDEWEB)

    Poupart, M. [Electricite de France, Paris (France)

    2004-09-01

    Electricite de France (EDF) recognizes that a dam's structural resistance and its capacity to withstand heavy flooding are 2 of the most significant safety aspects for hydroelectric power stations. However, in addition to dam failure, there are safety risks for the public who frequent the rivers up and down stream from the dam, as well as on property and the environment. A fatal accident which occurred in 1995 down river from EDF's Monteynard hydroelectric facility on the Drac River prompted the utility to take measures to improve control over this type of hazard. Collaboration with public authorities led to an action plan to educate the public about possible danger areas and to improve methods of reducing risks. Regulations regarding access to these areas were also studied along with ways of informing and warning the public. All the stretches of river directly above and below the power stations and dams were listed systematically and a qualitative analysis was carried out of any possible dangers. This led to changes in operating rules, technical instructions and power plant operating regulations. Certain areas are designated as prohibited areas, such as places subject to hazards caused by violent and unexpected water discharges. This paper outlined the Hydraulic Safety Significant Event (HSSE) classification which relates to an operational event related to water that is liable to affect human beings, the environment, water level or flow rate. 9 figs.

  5. 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)

  6. Environmental, Health and Safety Assessment: ATS 7H Program (Phase 3R) Test Activities at the GE Power Systems Gas Turbine Manufacturing Facility, Greenville, SC

    Energy Technology Data Exchange (ETDEWEB)

    None

    1998-11-17

    International Technology Corporation (IT) was contracted by General Electric Company (GE) to assist in the preparation of an Environmental, Health and Safety (HI&3) assessment of the implementation of Phase 3R of the Advanced Turbine System (ATS) 7H program at the GE Gas Turbines facility located in Greenville, South Carolina. The assessment was prepared in accordance with GE's contractual agreement with the U.S. Department of Energy (GE/DOE Cooperative Agreement DE-FC21-95MC3 1176) and supports compliance with the requirements of the National Environmental Policy Act of 1970. This report provides a summary of the EH&S review and includes the following: General description of current site operations and EH&S status, Description of proposed ATS 7H-related activities and discussion of the resulting environmental, health, safety and other impacts to the site and surrounding area. Listing of permits and/or licenses required to comply with federal, state and local regulations for proposed 7H-related activities. Assessment of adequacy of current and required permits, licenses, programs and/or plans.

  7. Overview of US fast-neutron facilities and testing capabilities

    International Nuclear Information System (INIS)

    Evans, E.A.; Cox, C.M.; Jackson, R.J.

    1982-01-01

    Rather than attempt a cataloging of the various fast neutron facilities developed and used in this country over the last 30 years, this paper will focus on those facilities which have been used to develop, proof test, and explore safety issues of fuels, materials and components for the breeder and fusion program. This survey paper will attempt to relate the evolution of facility capabilities with the evolution of development program which use the facilities. The work horse facilities for the breeder program are EBR-II, FFTF and TREAT. For the fusion program, RTNS-II and FMIT were selected

  8. 78 FR 48029 - Improving Chemical Facility Safety and Security

    Science.gov (United States)

    2013-08-07

    ... Improving Chemical Facility Safety and Security By the authority vested in me as President by the... at reducing the safety risks and security risks associated with hazardous chemicals. However... to further improve chemical facility safety and security in coordination with owners and operators...

  9. Safety of pulmonary function testing

    DEFF Research Database (Denmark)

    Roberts, Cara; Ward, Simon; Walsted, Emil

    2017-01-01

    BACKGROUND: Pulmonary function testing (PFT) is a key investigation in the evaluation of individuals with respiratory symptoms; however, the safety of routine and specialised PFT testing has not been reported in a large data set. Using patient safety incident (PSI) records, we aimed to assess risk...... was rated using the NHS National Patient Safety Agency and any hospital admission reported. RESULTS: There were 119 PSIs reported from 186 000 PFT; that is, 0.6 PSIs per 1000 tests. Cardiopulmonary PSIs were 3.3 times more likely to occur than non-cardiopulmonary (95% CI 2.17 to 5.12). Syncope was the most...

  10. Survey of solar thermal test facilities

    Energy Technology Data Exchange (ETDEWEB)

    Masterson, K.

    1979-08-01

    The facilities that are presently available for testing solar thermal energy collection and conversion systems are briefly described. Facilities that are known to meet ASHRAE standard 93-77 for testing flat-plate collectors are listed. The DOE programs and test needs for distributed concentrating collectors are identified. Existing and planned facilities that meet these needs are described and continued support for most of them is recommended. The needs and facilities that are suitable for testing components of central receiver systems, several of which are located overseas, are identified. The central contact point for obtaining additional details and test procedures for these facilities is the Solar Thermal Test Facilities Users' Association in Albuquerque, N.M. The appendices contain data sheets and tables which give additional details on the technical capabilities of each facility. Also included is the 1975 Aerospace Corporation report on test facilities that is frequently referenced in the present work.

  11. Drop test facility available to private industry

    International Nuclear Information System (INIS)

    Shappert, L.B.; Box, W.D.

    1983-01-01

    In 1978, a virtually unyielding drop test impact pad was constructed at Oak Ridge National Laboratory's (ORNL's) Tower Shielding Facility (TSF) for the testing of heavy shipping containers designed for transporting radioactive materials. Because of the facility's unique capability for drop-testing large, massive shipping packages, it has been identified as a facility which can be made available for non-DOE users

  12. Valve testing for UK PWR safety applications

    International Nuclear Information System (INIS)

    George, P.T.; Bryant, S.

    1989-01-01

    Extensive testing and development has been done by the Central Electricity Generating Board (CEGB) to support the design, construction and operation of Sizewell B, the UK's first PWR. A Blowdown Rig for the Assessment of Valve Operability - (BRAVO) has been constructed at the CEGB Marchwood Engineering Laboratory to reproduce PWR Pressurizer fluid conditions for the full scale testing of Pressurizer Relief System (PRS) valves. A full size tandem pair of Pilot Operated Safety Relief Valves (POSRVs) is being tested under the full range of pressurizer fluid conditions. Tests to date have produced important data on the performance of the valve in its Cold Overpressure protection mode of operation and on methods for the in-service testing of the valve. Also, a full size pressurizer safety valve has been tested under full PRS fluid conditions to develop a methodology for the pre-service testing of the Sizewell valves. Further work will be carried out to develop procedures for the in-service testing of the valve. In the Main Steam Safety Valve test program carried out at the Siemens-KWU Test Facilities, a single MSSV from three potential suppliers was tested under full secondary system conditions. The test results have been analyzed and are reflected in the CEGB's arrangements for the pre-service and in-service testing of the Sizewell MSSVs. Valves required to interrupt pipebreak flow must be qualified for this duty by testing or a combination of testing and analysis. To obtain guidance on the performance of such tests gate and globe valves have been subjected to simulated pipebreaks under PWR primary circuit conditions. In the light of problems encountered with gate valve closure under these conditions, further tests are currently being carried out on the BRAVO facility on a gate valve, in preparation for the full scale flow interruption qualification testing of the Sizewell main steam isolation valve

  13. The Brookhaven Accelerator Test Facility

    International Nuclear Information System (INIS)

    Batchelor, K.; Chou, T.S.; Fernow, R.C.; Fischer, J.; Gallardo, J.; Kirk, H.G.; Koul, R.; Palmer, R.B.; Pellegrini, C.; Sheehan, J.; Srinivasan-Rao, T.; Ulc, S.; Woodle, M.; Bigio, I.; Kurnit, N.; McDonald, K.T.

    1989-01-01

    The Brookhaven Accelerator Test Facility ATF will consist of a 50-100 MeV/c electron linac and a 100 GW CO 2 laser system. A high brightness RF-gun operating at 2,856 MHz is to be used as the injector into the linac. The RF-gun contains a Nd:Yag-laser-driven photocathode capable of producing a stream of six ps electron pulses separated by 12.5 ns. The maximum charge in a micropulse will be one nano-Coulomb. The CO 2 laser pulse length will be a few picoseconds and will be synchronized with the electron pulse. The first experimental beam is expected in Fall 89. The design electron beam parameters are given and possible initial experiments are discussed. 9 refs., 1 fig., 3 tabs

  14. The Brookhaven Accelerator Test Facility

    International Nuclear Information System (INIS)

    Batchelor, K.; Chou, T.S.; Fernow, R.C.

    1988-01-01

    The Brookhaven Accelerator Test Facility (ATF) will consist of a 50--100 MeV/c electron linac and a 100 GW CO 2 laser system. A high brightness RF-gun operating at 2856 MHz is to be used as the injector into the linac. The RF-gun contains a Nd:Yag-laser-driven photocathode capable of producing a stream of six ps electron pulses separated by 12.5 ns. The maximum charge in a micropulse will be one nano-Coulomb. The CO 2 laser pulse length will be a few picoseconds and will be synchronized with the electron pulse. The first experimental beam is expected in Fall 89. The design electron beam parameters are given and possible initial experiments are discussed. 9 refs., 1 fig., 3 tabs

  15. Criticality safety analysis for mockup facility

    International Nuclear Information System (INIS)

    Shin, Young Joon; Shin, Hee Sung; Kim, Ik Soo; Oh, Seung Chul; Ro, Seung Gy; Bae, Kang Mok

    2000-03-01

    Benchmark calculations for SCALE4.4 CSAS6 module have been performed for 31 UO 2 fuel, 15MOX fuel and 10 metal material criticality experiments and then calculation biases of the SCALE 4.4 CSAS6 module have been revealed to be 0.00982, 0.00579 and 0.02347, respectively. When CSAS6 is applied to the criticality safety analysis for the mockup facility in which several kinds of nuclear material components are included, the calculation bias of CSAS6 is conservatively taken to be 0.02347. With the aid of this benchmarked code system, criticality safety analyses for the mockup facility at normal and hypothetical accidental conditions have been carried out. It appears that the maximum K eff is 0.28356 well below than the critical limit, K eff =0.95 at normal condition. In a hypothetical accidental condition, the maximum K eff is found to be 0.73527 much lower than the subcritical limit. For another hypothetical accidental condition the nuclear material leaks out of container and spread or lump in the floor, it was assumed that the nuclear material is shaped into a slab and water exists in the empty space of the nuclear material. K eff has been calculated as function of slab thickness and the volume ratio of water to nuclear material. The result shows that the K eff increases as the water volume ratio increases. It is also revealed that the K eff reaches to the maximum value when water if filled in the empty space of nuclear material. The maximum K eff value is 0.93960 lower than the subcritical limit

  16. Validation testing of safety-critical software

    International Nuclear Information System (INIS)

    Kim, Hang Bae; Han, Jae Bok

    1995-01-01

    A software engineering process has been developed for the design of safety critical software for Wolsung 2/3/4 project to satisfy the requirements of the regulatory body. Among the process, this paper described the detail process of validation testing performed to ensure that the software with its hardware, developed by the design group, satisfies the requirements of the functional specification prepared by the independent functional group. To perform the tests, test facility and test software were developed and actual safety system computer was connected. Three kinds of test cases, i.e., functional test, performance test and self-check test, were programmed and run to verify each functional specifications. Test failures were feedback to the design group to revise the software and test results were analyzed and documented in the report to submit to the regulatory body. The test methodology and procedure were very efficient and satisfactory to perform the systematic and automatic test. The test results were also acceptable and successful to verify the software acts as specified in the program functional specification. This methodology can be applied to the validation of other safety-critical software. 2 figs., 2 tabs., 14 refs. (Author)

  17. Electromagnetic Interference (EMI) and TEMPEST Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Electromagnetic Interference (EMI), Electromagnetic Compatibility (EMC) and TEMPEST testing are conducted at EPG's Blacktail Canyon Test Facility in one of its two...

  18. 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)

  19. 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)

  20. Fuel Supply Shutdown Facility Interim Operational Safety Requirements

    International Nuclear Information System (INIS)

    BENECKE, M.W.

    2000-01-01

    The Interim Operational Safety Requirements for the Fuel Supply Shutdown (FSS) Facility define acceptable conditions, safe boundaries, bases thereof, and management of administrative controls to ensure safe operation of the facility

  1. Demonstration poloidal coil test facility

    International Nuclear Information System (INIS)

    Sato, Masahiko; Kawano, Katumi; Tada, Eisuke

    1989-01-01

    A new compact cryogenic cold compressor was developed by Japan Atomic Energy Research Institute (JAERI) in collaboration with Isikawajima-Harima Heavy Industries Co., Ltd. (IHI) in order to produce the supercritical helium below 4.2 K for Demonstration Poloidal Coils (DPC) which are forced-flow cooled type superconducting pulse coils. This compressor is one of key components for DPC test facility. The cold compressor reduces pressure in liquid helium bath, which contains liquid helium of around 3,000 l, down to 0.5 atm efficiently. Consequently, supercritical helium down to 3.5 K is produced and supplied to the DPC coils. A centrifugal compressor with dynamic gas bearing is selected as a compressor mechanism to realize high adiabatic efficiency and large flow rate. In this performance tests, the compressor was operated for 220 h at saturated condition from 0.5 to 1.0 atm without any failure. High adiabatic efficiency (more than 60 %) is achieved with wide flow range (25-65 g/s) and the design value is fully satisfied. The compressor can rotate up to 80,000 rpm at maximum then the coil supply temperature of supercritical helium is 3.5 K. (author)

  2. SRL incinerator components test facility

    International Nuclear Information System (INIS)

    Freed, E.J.

    1982-08-01

    A full-scale (5 kg waste/hour) controlled-air incinerator, the ICTF, is presently being tested with simulated waste as part of a program to develop technology for incineration of Savannah River Plant solid transuranic wastes. This unit is designed specifically to incinerate relatively small quantities of solid combustible waste that are contaminated up to 10 5 times the present nominal 10 nCi/g threshold value for such isotopes as 238 Pu, 239 Pu, 242 Cm, and 252 Cf. Automatic incinerator operation and control has been incorporated into the design, simulating the future plant design which minimizes operator radiation exposure. Over 3000 kg of nonradioactive wastes characteristic of plutonium finishing operations have been incinerated at throughputs exceeding 5 kg/hr. Safety and reliability were the major design objectives. In addition to the incinerator tests, technical data were gathered on two different off-gas systems: a wet system composed of three scrubbers in series, and a dry system employing sintered metal filters

  3. Successful start for new CLIC test facility

    CERN Document Server

    2004-01-01

    A new test facility is being built to study key feasibility issues for a possible future linear collider called CLIC. Commissioning of the first part of the facility began in June 2003 and nominal beam parameters have been achieved already.

  4. 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.

  5. 77 FR 2606 - Pipeline Safety: Random Drug Testing Rate

    Science.gov (United States)

    2012-01-18

    ... DEPARTMENT OF TRANSPORTATION Pipeline and Hazardous Materials Safety Administration [Docket ID PHMSA-2012-0004] Pipeline Safety: Random Drug Testing Rate AGENCY: Pipeline and Hazardous Materials... pipelines and operators of liquefied natural gas facilities must select and test a percentage of covered...

  6. 75 FR 9018 - Pipeline Safety: Random Drug Testing Rate

    Science.gov (United States)

    2010-02-26

    ... DEPARTMENT OF TRANSPORTATION Pipeline and Hazardous Materials Safety Administration [Docket ID PHMSA-2010-0034] Pipeline Safety: Random Drug Testing Rate AGENCY: Pipeline and Hazardous Materials... pipelines and operators of liquefied natural gas facilities must select and test a percentage of covered...

  7. Field Lysimeter Test Facility for protective barriers: Experimental plan

    International Nuclear Information System (INIS)

    Kirkham, R.R.; Gee, G.W.; Downs, J.L.

    1987-12-01

    This document was first written in October 1986 and has been used to guide the design of the Field Lysimeter Test Facility (FLTF) and to promote discussions between research and engineering staff regarding the selection of barrier treatments for inclusion in the FLTF. The construction of the lysimeter facility was completed June 28, 1987. This document describes the facility, the treatments placed in each lysimeter, types of measurements made in each lysimeter, and a brief discussion of project activities related to quality assurance, safety, and funding requirements. The treatment description and figures have been updated to reflect the lysimeter facility as constructed. 12 refs., 6 figs., 5 tabs

  8. Cryogenic test facility at VECC, Kolkata

    International Nuclear Information System (INIS)

    Sarkar, Amit; Bhunia, Uttam; Pradhan, J.; Sur, A.; Bhandari, R.K.; Ranganathan, R.

    2003-01-01

    In view of proposed K-500 superconducting cyclotron project, cryogenic test facility has been set up at the centre. The facility can broadly be categorized into two- a small scale test facility and a large scale test facility. This facility has been utilized for the calibration of liquid helium level probe, cryogenic temperature probe, and I-B plot for a 7 T superconducting magnet. Spiral-shaped superconducting short sample with specific dimension and specially designed stainless steel sample holder has already been developed for the electrical characterisation. The 1/5 th model superconducting coil along with its quench detection circuit and dump resistor has also been developed

  9. Activity of safety review for the facilities using nuclear material (2). Safety review results and maintenance experiences for hot laboratories

    International Nuclear Information System (INIS)

    Amagai, Tomio; Fujishima, Tadatsune; Mizukoshi, Yasutaka; Sakamoto, Naoki; Ohmori, Tsuyoshi

    2009-01-01

    In the site of O-arai Research and Development Center of Japan Atomic Energy Agency (JAEA), five hot laboratories for post-irradiation examination and development of plutonium fuels are operated more than 30 years. A safety review method for preventive maintenance on these hot laboratories includes test facilities and devices are established in 2003. After that, the safety review of these facilities and devices are done and taken the necessary maintenance based on the results in each year. In 2008, 372 test facilities and devices in these hot laboratories were checked and reviewed by this method. As a results of the safety review, repair issues of 38 facilities of above 372 facilities were resolved. This report shows the review results and maintenance experiences based on the results. (author)

  10. Safety analysis of DUPIC fuel development facility

    International Nuclear Information System (INIS)

    Lee, H. H.; Park, J. J.; Shin, J. M.; Yang, M. S.; Baek, S. Y.; Ahn, J. Y.

    2001-01-01

    Various experimental facilities are necessary in order to perform experimental verification for development of DUPIC fuel fabrication technology. In special, since highly radioactive material such as spent PWR fuel is used for this experiment, DUPIC fuel fabrication has to be performed in hot cell by remote handling. Therefore, it should be provided with proper engineering requirement and safety. M6 hot cell of IMEF which is to used for DUPIC fuel fabrication experiment was constructed as an α-γ hot cell for material examination of small amount of high-burnup fuel. The characteristics and amount of spent fuel for DUPIC fuel fabrication experiment will be different from the original design criteria. Therefore, the increased amount of spent fuel and different characteristics of experiment result in not only change of shielding and enviornmental evaluation results but new requirement of nuclear criticality evaluation. Therefore, this study includes evaluation of shielding, environmental effect and nuclear criticality in case that IMEF M6 hot cell is used for DUPIC fuel fabrication

  11. 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

  12. Argonne's new Wakefield Test Facility

    International Nuclear Information System (INIS)

    Simpson, J.D.

    1992-01-01

    The first phase of a high current, short bunch length electron beam research facility, the AWA, is near completion at Argonne. At the heart of the facility is a photocathode based electron gun and accelerating sections designed to deliver 20 MeV pulses with up to 100 nC per pulse and with pulse lengths of approximately 15 ps (fw). Using a technique similar to that originated at Argonne's AATF facility, a separate weak probe pulse can be generated and used to diagnose wake effects produced by the intense pulses. Initial planned experiments include studies of plasma wakefields and dielectric wakefield devices, and expect to demonstrate large, useful accelerating gradients (> 100 MeV/m). Later phases of the facility will increase the drive bunch energy to more than 100 MeV to enable acceleration experiments up to the GeV range. Specifications, design details, and commissioning progress are presented

  13. 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

  14. Fast Flux Test Facility (FFTF) maintenance provisions

    International Nuclear Information System (INIS)

    Marshall, J.L.

    1981-05-01

    The Fast Flux Test Facility (FFTF) was designed with maintainability as a primary parameter, and facilities and provisions were designed into the plant to accommodate the maintenance function. This paper describes the FFTF and its systems. Special maintenance equipment and facilities for performing maintenance on radioactive components are discussed. Maintenance provisions designed into the plant to enhance maintainability are also described

  15. Controlled Archaeological Test Site (CATS) Facility

    Data.gov (United States)

    Federal Laboratory Consortium — CATS facility is at the Construction Engineering Research Laboratory (CERL), Champaign, IL. This 1-acre test site includes a variety of subsurface features carefully...

  16. Safety analysis report for the Waste Storage Facility. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Bengston, S.J.

    1994-05-01

    This safety analysis report outlines the safety concerns associated with the Waste Storage Facility located in the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. The three main objectives of the report are: define and document a safety basis for the Waste Storage Facility activities; demonstrate how the activities will be carried out to adequately protect the workers, public, and environment; and provide a basis for review and acceptance of the identified risk that the managers, operators, and owners will assume.

  17. Facts about food irradiation: Safety of irradiation facilities

    International Nuclear Information System (INIS)

    1991-01-01

    This fact sheet considers the safety of industrial irradiation facilities. Although there have been accidents, none of them has endangered public health or environmental safety, and the radiation processing industry is considered to have a very good safety record. Gamma irradiators do not produce radioactive waste, and the radiation sources at the facilities cannot explode nor in any other way release radioactivity into the environment. 3 refs

  18. X-ray and nuclear radiation facilities: personnel safety features

    International Nuclear Information System (INIS)

    Mason, W.J.; Pipes, E.W.; Rucker, T.R.; Smith, D.N.; West, C.M.

    1976-10-01

    The Oak Ridge Y-12 Plant is a research and production installation. The nature and versatility of this work require the use of a large number and variety of x-ray and radiographic sources for nondestructive testing and material analyses. Presently, there are over 80 x-ray generators in the plant, which range in size from small, portable units which operate at a less than 50 kilovolts potential and 0.1 milliampere current to an electron linear accelerator which operates at 12-million electron volts and produces a radiation beam of such intensity that it could deliver a lethal dose to man in a fraction of a minute. There are also almost 50 gamma and neutron sources in use in the plant. These units range in size from a few millicuries to several hundred curies. Although the radiation safety at each of these facilities was considered adequate, the administrative and maintenance procedures became unduly complicated. Accordingly, engineering standards and uniform operating procedures were considered necessary to alleviate these complications and, in so doing, provide an improved measure of radiation safety. Development and implementation of these standards are described and the general philosophy and approach to these standards are outlined. Use of a matrix (type of installation versus radiation safety feature) to facilitate equipment classification and personnel safety feature requirements is presented. Included is a set of the standards showing formats, matrices, etc., and the detailed standards for each safety feature

  19. 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

  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. Supervision of the safety culture in nuclear facilities

    International Nuclear Information System (INIS)

    2014-11-01

    This brochure issued by the Swiss Federal Nuclear Safety Inspectorate ENSI reports on safety culture aspects in nuclear facilities and ENSI’s activities as a supervisory instance. ENSI is the independent supervisory authority for the nuclear sector in Switzerland. A definition of safety culture is presented and the development of the concepts used in its monitoring are discussed. The main attributes of a good safety culture are discussed. Further, the conceptual basics and principles of such monitoring are looked at and the methods used for the supervision of safety culture in nuclear facilities are described

  2. 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)

  3. Engineered Barrier Test Facility status report, 1984

    International Nuclear Information System (INIS)

    Phillips, S.J.; Adams, M.R.; Gilbert, T.W.; Meinhardt, C.C.; Mitchell, R.M.; Waugh, W.J.

    1985-02-01

    This report provides a general summary of activities completed to date at the Hanford Engineered Barrier Test Facility. This facility is used to test and compare construction practices and performance of alternative designs of engineered barrier cover systems. These cover systems are being evaluated for potential use for isolation and confinement of buried waste disposal structures

  4. Preparation of safety and regulatory document for BARC Facilities

    International Nuclear Information System (INIS)

    Prasad, S.S.; Jayarajan, K.

    2017-01-01

    In India, the necessary codes and safety guidelines for achieving the safety objectives are provided by the Atomic Energy Regulatory Board (AERB), which are in conformity with the principles of radiation protection as formulated by the International Council of Radiation Protection (ICRP) and International Atomic Energy Agency (IAEA). The same is followed by BARC Safety Council (BSC), which is the regulatory body for the BARC facilities. In addition to all types of fuel cycle facilities, BSC regulates safety of many types of conventional facilities. Many such types of facilities and projects are not under the regulatory purview of AERB. Therefore, the Council has also initiated a programme for development and publication of safety documents for installations in BARC in the fields/ topics yet not addressed by IAEA or AERB. This makes the task pioneering, as some of the areas taken up for defining the regulatory requirements are new, where standard regulatory documents are not available

  5. Scaling analysis for the OSU AP600 test facility (APEX)

    International Nuclear Information System (INIS)

    Reyes, J.N.

    1998-01-01

    In this paper, the authors summarize the key aspects of a state-of-the-art scaling analysis (Reyes et al. (1995)) performed to establish the facility design and test conditions for the advanced plant experiment (APEX) at Oregon State University (OSU). This scaling analysis represents the first, and most comprehensive, application of the hierarchical two-tiered scaling (H2TS) methodology (Zuber (1991)) in the design of an integral system test facility. The APEX test facility, designed and constructed on the basis of this scaling analysis, is the most accurate geometric representation of a Westinghouse AP600 nuclear steam supply system. The OSU APEX test facility has served to develop an essential component of the integral system database used to assess the AP600 thermal hydraulic safety analysis computer codes. (orig.)

  6. Safety evaluation report of hot cell facilities for demonstration of advanced spent fuel conditioning process

    International Nuclear Information System (INIS)

    You, Gil Sung; Choung, W. M.; Ku, J. H.; Cho, I. J.; Kook, D. H.; Park, S. W.; Bek, S. Y.; Lee, E. P.

    2004-10-01

    The advanced spent fuel conditioning process(ACP) proposed to reduce the overall volume of the PWR spent fuel and improve safety and economy of the long-term storage of spent fuel. In the next phase(2004∼2006), the hot test will be carried out for verification of the ACP in a laboratory scale. For the hot test, the hot cell facilities of α- type and auxiliary facilities are required essentially for safe handling of high radioactive materials. As the hot cell facilities for demonstration of the ACP, a existing hot cell of β- type will be refurbished to minimize construction expenditures of hot cell facility. Up to now, the detail design of hot cell facilities and process were completed, and the safety analysis was performed to substantiate secure of conservative safety. The design data were submitted for licensing which was necessary for construction and operation of hot cell facilities. The safety investigation of KINS on hot cell facilities was completed, and the license for construction and operation of hot cell facilities was acquired already from MOST. In this report, the safety analysis report submitted to KINS was summarized. And also, the questionnaires issued from KINS and answers of KAERI in process of safety investigation were described in detail

  7. Energy Systems Test Area (ESTA). Power Systems Test Facilities

    Science.gov (United States)

    Situ, Cindy H.

    2010-01-01

    This viewgraph presentation provides a detailed description of the Johnson Space Center's Power Systems Facility located in the Energy Systems Test Area (ESTA). Facilities and the resources used to support power and battery systems testing are also shown. The contents include: 1) Power Testing; 2) Power Test Equipment Capabilities Summary; 3) Source/Load; 4) Battery Facilities; 5) Battery Test Equipment Capabilities Summary; 6) Battery Testing; 7) Performance Test Equipment; 8) Battery Test Environments; 9) Battery Abuse Chambers; 10) Battery Abuse Capabilities; and 11) Battery Test Area Resources.

  8. Design aspects of radiological safety in nuclear facilities

    International Nuclear Information System (INIS)

    Patkulkar, D.S.; Purohit, R.G.; Tripathi, R.M.

    2014-01-01

    In order to keep operational performance of a nuclear facility high and to keep occupational and public exposure ALARA, radiological safety provisions must be reviewed at the time of facility design. Deficiency in design culminates in deteriorated system performance and non adherence to safety standards and could sometimes result in radiological incident. Important radiological aspects relevant to safety were compiled based on operating experiences, design deficiencies brought out from past nuclear incidents, experience gained during maintenance, participation in design review of upcoming nuclear facilities and radiological emergency preparedness

  9. Criticality Safety Evaluation of Hanford Tank Farms Facility

    Energy Technology Data Exchange (ETDEWEB)

    WEISS, E.V.

    2000-12-15

    Data and calculations from previous criticality safety evaluations and analyses were used to evaluate criticality safety for the entire Tank Farms facility to support the continued waste storage mission. This criticality safety evaluation concludes that a criticality accident at the Tank Farms facility is an incredible event due to the existing form (chemistry) and distribution (neutron absorbers) of tank waste. Limits and controls for receipt of waste from other facilities and maintenance of tank waste condition are set forth to maintain the margin subcriticality in tank waste.

  10. Criticality Safety Evaluation of Hanford Tank Farms Facility

    International Nuclear Information System (INIS)

    WEISS, E.V.

    2000-01-01

    Data and calculations from previous criticality safety evaluations and analyses were used to evaluate criticality safety for the entire Tank Farms facility to support the continued waste storage mission. This criticality safety evaluation concludes that a criticality accident at the Tank Farms facility is an incredible event due to the existing form (chemistry) and distribution (neutron absorbers) of tank waste. Limits and controls for receipt of waste from other facilities and maintenance of tank waste condition are set forth to maintain the margin subcriticality in tank waste

  11. 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

  12. Safety distance between underground natural gas and water pipeline facilities

    International Nuclear Information System (INIS)

    Mohsin, R.; Majid, Z.A.; Yusof, M.Z.

    2014-01-01

    A leaking water pipe bursting high pressure water jet in the soil will create slurry erosion which will eventually erode the adjacent natural gas pipe, thus causing its failure. The standard 300 mm safety distance used to place natural gas pipe away from water pipeline facilities needs to be reviewed to consider accidental damage and provide safety cushion to the natural gas pipe. This paper presents a study on underground natural gas pipeline safety distance via experimental and numerical approaches. The pressure–distance characteristic curve obtained from this experimental study showed that the pressure was inversely proportional to the square of the separation distance. Experimental testing using water-to-water pipeline system environment was used to represent the worst case environment, and could be used as a guide to estimate appropriate safety distance. Dynamic pressures obtained from the experimental measurement and simulation prediction mutually agreed along the high-pressure water jetting path. From the experimental and simulation exercises, zero effect distance for water-to-water medium was obtained at an estimated horizontal distance at a minimum of 1500 mm, while for the water-to-sand medium, the distance was estimated at a minimum of 1200 mm. - Highlights: • Safe separation distance of underground natural gas pipes was determined. • Pressure curve is inversely proportional to separation distance. • Water-to-water system represents the worst case environment. • Measured dynamic pressures mutually agreed with simulation results. • Safe separation distance of more than 1200 mm should be applied

  13. Design of concrete structures important to safety of nuclear facilities

    International Nuclear Information System (INIS)

    2001-10-01

    Civil engineering structures in nuclear installations form an important feature having implications to safety performance of these installations. The objective and minimum requirements for the design of civil engineering buildings/structures to be fulfilled to provide adequate assurance for safety of nuclear installations in India (such as pressurised heavy water reactor and related systems) are specified in the Safety standard for civil engineering structures important to safety of nuclear facilities. This standard is written by AERB to specify guidelines for implementation of the above civil engineering safety standard in the design of concrete structures important to safety

  14. CryoModule Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — CMTFis able to test complete SRF cryomodules at cryogenic operating temperatures and with RF Power. CMTF will house the PIP-II Injector Experiment allowing test of...

  15. Naval Aerodynamics Test Facility (NATF)

    Data.gov (United States)

    Federal Laboratory Consortium — The NATF specializes in Aerodynamics testing of scaled and fullsized Naval models, research into flow physics found on US Navy planes and ships, aerosol testing and...

  16. Enhancement of safety for reprocessing facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-08-15

    After the accident in Fukushima Daiichi Nuclear Power Station, eight emergency projects taking into account the accident were newly launched in JNES. This project for a reprocessing facility was one of them. Major items conducted in the project were as follows. (1) Researches, studies and evaluations etc. on various events under a total AC (alternating current) power loss condition Under this condition following subjects of the events were performed. a) An equipment with a removing function of decay heat and a time to reach a certain critical condition, e.g. a solution boiling, b) An equipment with a preventing function of accumulation of hydrogen gas and a time to reach a concentration of hydrogen gas to that of the lowest limit of combustion, c) Specifications of an alternative electric source and how to supply power. (2) Researches, studies and evaluations etc. on beyond design basis events. Following subjects on these events were performed. a) An event progression scenario, a consequence, a time period between an initiating event and a resultant accident or a certain critical condition, and draft inspection criteria, b) Draft inspection criteria for a stress test. (author)

  17. Explotation of irradiation facilities. Safety handbook

    International Nuclear Information System (INIS)

    Prieto Miranda, Enrique Franscisco; Melo Crespo, Jose Carlos

    1997-01-01

    At present in the world there are more of 160 gamma radiation facilities and more of 600 electron bean accelerators in operation, at least one in each member state of International Atomic Energy Agency. In this paper is elaborated a Manual with the security criteria to operation of these facility types

  18. 340 Waste handling Facility Hazard Categorization and Safety Analysis

    International Nuclear Information System (INIS)

    Rodovsky, T.J.

    2010-01-01

    The analysis presented in this document provides the basis for categorizing the facility as less than Hazard Category 3. The final hazard categorization for the deactivated 340 Waste Handling Facility (340 Facility) is presented in this document. This hazard categorization was prepared in accordance with DOE-STD-1 027-92, Change Notice 1, Hazard Categorization and Accident Analysis Techniques for Compliance with Doe Order 5480.23, Nuclear Safety Analysis Reports. The analysis presented in this document provides the basis for categorizing the facility as less than Hazard Category (HC) 3. Routine nuclear waste receiving, storage, handling, and shipping operations at the 340 Facility have been deactivated, however, the facility contains a small amount of radioactive liquid and/or dry saltcake in two underground vault tanks. A seismic event and hydrogen deflagration were selected as bounding accidents. The generation of hydrogen in the vault tanks without active ventilation was determined to achieve a steady state volume of 0.33%, which is significantly less than the lower flammability limit of 4%. Therefore, a hydrogen deflagration is not possible in these tanks. The unmitigated release from a seismic event was used to categorize the facility consistent with the process defined in Nuclear Safety Technical Position (NSTP) 2002-2. The final sum-of-fractions calculation concluded that the facility is less than HC 3. The analysis did not identify any required engineered controls or design features. The Administrative Controls that were derived from the analysis are: (1) radiological inventory control, (2) facility change control, and (3) Safety Management Programs (SMPs). The facility configuration and radiological inventory shall be controlled to ensure that the assumptions in the analysis remain valid. The facility commitment to SMPs protects the integrity of the facility and environment by ensuring training, emergency response, and radiation protection. The full scale

  19. Construction and commissioning test report of the CEDM test facility

    Energy Technology Data Exchange (ETDEWEB)

    Chung, C. H.; Kim, J. T.; Park, W. M.; Youn, Y. J.; Jun, H. G.; Choi, N. H.; Park, J. K.; Song, C. H.; Lee, S. H.; Park, J. K

    2001-02-01

    The test facility for performance verification of the control element drive mechanism (CEDM) of next generation power plant was installed at the site of KAERI. The CEDM was featured a mechanism consisting of complicated mechanical parts and electromagnetic control system. Thus, a new CEDM design should go through performance verification tests prior to it's application in a reactor. The test facility can simulate the reactor operating conditions such as temperature, pressure and water quality and is equipped with a test chamber to accomodate a CEDM as installed in the power plant. This test facility can be used for the following tests; endurance test, coil cooling test, power measurement and reactivity rod drop test. The commissioning tests for the test facility were performed up to the CEDM test conditions of 320 C and 150 bar, and required water chemistry was obtained by operating the on-line water treatment system.

  20. Construction and commissioning test report of the CEDM test facility

    International Nuclear Information System (INIS)

    Chung, C. H.; Kim, J. T.; Park, W. M.; Youn, Y. J.; Jun, H. G.; Choi, N. H.; Park, J. K.; Song, C. H.; Lee, S. H.; Park, J. K.

    2001-02-01

    The test facility for performance verification of the control element drive mechanism (CEDM) of next generation power plant was installed at the site of KAERI. The CEDM was featured a mechanism consisting of complicated mechanical parts and electromagnetic control system. Thus, a new CEDM design should go through performance verification tests prior to it's application in a reactor. The test facility can simulate the reactor operating conditions such as temperature, pressure and water quality and is equipped with a test chamber to accomodate a CEDM as installed in the power plant. This test facility can be used for the following tests; endurance test, coil cooling test, power measurement and reactivity rod drop test. The commissioning tests for the test facility were performed up to the CEDM test conditions of 320 C and 150 bar, and required water chemistry was obtained by operating the on-line water treatment system

  1. The construction of solid waste form test facility

    International Nuclear Information System (INIS)

    Park, Hun Hwee; Kim, Joon Hyung; Lee, Byung Jik; Koo, Jun Mo; Kim, Jeong Guk; Jung, In Ha

    1990-03-01

    The solid waste form test facility (SWFTF) to test and/or evaluate the characteristics of waste forms, such as homogeniety, mechanical properties, thermal properties, waste resistance and leachability, have been constructed, and some equipments for testing actual waste forms has been purchased; radiocative monitoring system, glove box for the manipulator repair room, and uninteruppted power supply system, et al. Classifications of radioactive wastes, basic requirements and criteria to be considered during waste management were also reviewed. Some of the described items above have been standardized for the purpose of indigenigation. Therefore, safety assurance of waste forms, as well as increase in the range of participating of domestic companies in construction of further nuclear facilities could be obtained as results through constructing this facility. In the furture this facility is going to be utilized not only for the inspection of waste forms but also for the periodic decontamination for extending the life time of some expensive radiological equipments using remote handling techniques. (author)

  2. 78 FR 69433 - Executive Order 13650 Improving Chemical Facility Safety and Security Listening Sessions

    Science.gov (United States)

    2013-11-19

    ... Chemical Facility Safety and Security Listening Sessions AGENCY: National Protection and Programs... from stakeholders on issues pertaining to Improving Chemical Facility Safety and Security (Executive... regulations, guidance, and policies; and identifying best practices in chemical facility safety and security...

  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. Battery Post-Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — Post-test diagnostics of aged batteries can provide additional information regarding the cause of performance degradation, which, previously, could be only inferred...

  5. Ballast Water Treatment Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — FUNCTION: Provides functionality for the full-scale testing and controlled simulation of ship ballasting operations for assessment of aquatic nuisance species (ANS)...

  6. Radiation Safety of Accelerator Facility with Regard to Regulation

    International Nuclear Information System (INIS)

    Dedi Sunaryadi; Gloria Doloresa

    2003-01-01

    The radiation safety of accelerator facility and the status of the facilities according to licensee in Indonesia as well as lesson learned from the accidents are described. The atomic energy Act No. 10 of 1997 enacted by the Government of Indonesia which is implemented in Radiation Safety Government Regulation No. 63 and 64 as well as practice-specific model regulation for licensing request are discussed. (author)

  7. Testing lifting systems in nuclear facilities

    International Nuclear Information System (INIS)

    Kling, H.; Laug, R.

    1984-01-01

    Lifting systems in nuclear facilities must be inspected at regular intervals after having undergone their first acceptance test. These inspections are frequently carried out by service firms which not only employ the skilled personnel required for such jobs but also make available the necessary test equipment. The inspections in particular include a number of sophisticated load tests for which test load systems have been developed to allow lifting systems to be tested so that reactor specific boundary conditions are taken into account. In view of the large number of facilities to be inspected, the test load system is a modular system. (orig.) [de

  8. E-4 Test Facility Design Status

    Science.gov (United States)

    Ryan, Harry; Canady, Randy; Sewell, Dale; Rahman, Shamim; Gilbrech, Rick

    2001-01-01

    Combined-cycle propulsion technology is a strong candidate for meeting NASA space transportation goals. Extensive ground testing of integrated air-breathing/rocket system (e.g., components, subsystems and engine systems) across all propulsion operational modes (e.g., ramjet, scramjet) will be needed to demonstrate this propulsion technology. Ground testing will occur at various test centers based on each center's expertise. Testing at the NASA John C. Stennis Space Center will be primarily concentrated on combined-cycle power pack and engine systems at sea level conditions at a dedicated test facility, E-4. This paper highlights the status of the SSC E-4 test Facility design.

  9. Design and Testing of BACRA, a Web-Based Tool for Middle Managers at Health Care Facilities to Lead the Search for Solutions to Patient Safety Incidents

    Science.gov (United States)

    Mira, José Joaquín; Vicente, Maria Asuncion; Fernandez, Cesar; Guilabert, Mercedes; Ferrús, Lena; Zavala, Elena; Silvestre, Carmen; Pérez-Pérez, Pastora

    2016-01-01

    Background Lack of time, lack of familiarity with root cause analysis, or suspicion that the reporting may result in negative consequences hinder involvement in the analysis of safety incidents and the search for preventive actions that can improve patient safety. Objective The aim was develop a tool that enables hospitals and primary care professionals to immediately analyze the causes of incidents and to propose and implement measures intended to prevent their recurrence. Methods The design of the Web-based tool (BACRA) considered research on the barriers for reporting, review of incident analysis tools, and the experience of eight managers from the field of patient safety. BACRA’s design was improved in successive versions (BACRA v1.1 and BACRA v1.2) based on feedback from 86 middle managers. BACRA v1.1 was used by 13 frontline professionals to analyze incidents of safety; 59 professionals used BACRA v1.2 and assessed the respective usefulness and ease of use of both versions. Results BACRA contains seven tabs that guide the user through the process of analyzing a safety incident and proposing preventive actions for similar future incidents. BACRA does not identify the person completing each analysis since the password introduced to hide said analysis only is linked to the information concerning the incident and not to any personal data. The tool was used by 72 professionals from hospitals and primary care centers. BACRA v1.2 was assessed more favorably than BACRA v1.1, both in terms of its usefulness (z=2.2, P=.03) and its ease of use (z=3.0, P=.003). Conclusions BACRA helps to analyze incidents of safety and to propose preventive actions. BACRA guarantees anonymity of the analysis and reduces the reluctance of professionals to carry out this task. BACRA is useful and easy to use. PMID:27678308

  10. Design and Testing of BACRA, a Web-Based Tool for Middle Managers at Health Care Facilities to Lead the Search for Solutions to Patient Safety Incidents.

    Science.gov (United States)

    Carrillo, Irene; Mira, José Joaquín; Vicente, Maria Asuncion; Fernandez, Cesar; Guilabert, Mercedes; Ferrús, Lena; Zavala, Elena; Silvestre, Carmen; Pérez-Pérez, Pastora

    2016-09-27

    Lack of time, lack of familiarity with root cause analysis, or suspicion that the reporting may result in negative consequences hinder involvement in the analysis of safety incidents and the search for preventive actions that can improve patient safety. The aim was develop a tool that enables hospitals and primary care professionals to immediately analyze the causes of incidents and to propose and implement measures intended to prevent their recurrence. The design of the Web-based tool (BACRA) considered research on the barriers for reporting, review of incident analysis tools, and the experience of eight managers from the field of patient safety. BACRA's design was improved in successive versions (BACRA v1.1 and BACRA v1.2) based on feedback from 86 middle managers. BACRA v1.1 was used by 13 frontline professionals to analyze incidents of safety; 59 professionals used BACRA v1.2 and assessed the respective usefulness and ease of use of both versions. BACRA contains seven tabs that guide the user through the process of analyzing a safety incident and proposing preventive actions for similar future incidents. BACRA does not identify the person completing each analysis since the password introduced to hide said analysis only is linked to the information concerning the incident and not to any personal data. The tool was used by 72 professionals from hospitals and primary care centers. BACRA v1.2 was assessed more favorably than BACRA v1.1, both in terms of its usefulness (z=2.2, P=.03) and its ease of use (z=3.0, P=.003). BACRA helps to analyze incidents of safety and to propose preventive actions. BACRA guarantees anonymity of the analysis and reduces the reluctance of professionals to carry out this task. BACRA is useful and easy to use.

  11. Final design of ITER port plug test facility

    Energy Technology Data Exchange (ETDEWEB)

    Cerisier, Thierry, E-mail: thierry.cerisier@yahoo.fr [ITER Organization, Route de Vinon-sur-Verdon, CS 90046, St Paul-lez-Durance Cedex, 13067 (France); Levesy, Bruno [ITER Organization, Route de Vinon-sur-Verdon, CS 90046, St Paul-lez-Durance Cedex, 13067 (France); Romannikov, Alexander [Institution “Project Center ITER”, Kurchatov sq. 1, Building 3, Moscow 123182 (Russian Federation); Rumyantsev, Yuri [JSC “Cryogenmash”, Moscow reg., Balashikha 143907 (Russian Federation); Cordier, Jean-Jacques; Dammann, Alexis [ITER Organization, Route de Vinon-sur-Verdon, CS 90046, St Paul-lez-Durance Cedex, 13067 (France); Minakov, Victor; Rosales, Natalya; Mitrofanova, Elena [JSC “Cryogenmash”, Moscow reg., Balashikha 143907 (Russian Federation); Portone, Sergey; Mironova, Ekaterina [Institution “Project Center ITER”, Kurchatov sq. 1, Building 3, Moscow 123182 (Russian Federation)

    2016-11-01

    Highlights: • We introduce the port plug test facility (purpose and status of the design). • We present the PPTF sub-systems. • We present the environmental and functional tests. • We present the occupational and nuclear safety functions. • We conclude on the achievements and next steps. - Abstract: To achieve the overall ITER machine availability target, the availability of diagnostics and heating port plugs shall be as high as 99.5%. To fulfill this requirement, it is mandatory to test the port plugs at operating temperature before installation on the machine and after refurbishment. The ITER port plug test facility (PPTF) is composed of several test stands that can be used to test the port plugs whereas at the end of manufacturing (in a non-nuclear environment), or after refurbishment in the ITER hot cell facility. The PPTF provides the possibility to perform environmental (leak tightness, vacuum and thermo-hydraulic performances) and functional tests (radio frequency acceptance tests, behavior of the plugs’ steering mechanism and calibration of diagnostics) on upper and equatorial port plugs. The final design of the port plug test facility is described. The configuration of the standalone test stands and the integration in the hot cell facility are presented.

  12. The PANDA facility and first test results

    International Nuclear Information System (INIS)

    Dreier, J.; Huggenberger, M.; Aubert, C.; Bandurski, T.; Fischer, O.; Healzer, J.; Lomperski, S.; Strassberger, H.J.; Varadi, G.; Yadigaroglu, G.

    1996-01-01

    The PANDA test facility at the Paul Scherrer Institute is used to study the long-term performance of the Simplified Boiling Water Reactor's passive containment cooling system. The PANDA tests demonstrate performance on a larger scale than previous tests and examine the effects of any non-uniform spatial distributions of steam and non-condensable gases in the system. The facility is in 1:1 vertical scale and 1:25 scale for volume, power etc. Extensive facility characterization tests and steady-state passive containment condenser performance tests are presented. The results of the base case test of a series of transient system behaviour tests are reviewed. The first PANDA tests exhibited reproducibility, and indicated that the Simplified Boiling Water Reactor's containment is likely to be favorably responsive and highly robust to changes in the thermal transport patterns. (orig.) [de

  13. Fast flux test facility hazards assessment

    International Nuclear Information System (INIS)

    Sutton, L.N.

    1994-01-01

    This document establishes the technical basis in support of Emergency Planning Activities for the Fast Flux Test Facility on the Hanford Site. The document represents an acceptable interpretation of the implementing guidance document for DOE Order 5500.3A. Through this document, the technical basis for the development of facility specific Emergency Action Levels and the Emergency Planning Zone is demonstrated

  14. A security/safety survey of long term care facilities.

    Science.gov (United States)

    Acorn, Jonathan R

    2010-01-01

    What are the major security/safety problems of long term care facilities? What steps are being taken by some facilities to mitigate such problems? Answers to these questions can be found in a survey of IAHSS members involved in long term care security conducted for the IAHSS Long Term Care Security Task Force. The survey, the author points out, focuses primarily on long term care facilities operated by hospitals and health systems. However, he believes, it does accurately reflect the security problems most long term facilities face, and presents valuable information on security systems and practices which should be also considered by independent and chain operated facilities.

  15. Safety of magnetic fusion facilities: Volume 2, Guidance

    International Nuclear Information System (INIS)

    1995-01-01

    This document provides guidance for the implementation of the requirements identified in Vol. 1 of this Standard. This guidance is intended for the managers, designers, operators, and other personnel with safety responsibilities for facilities designated as magnetic fusion facilities. While Vol. 1 is generally applicable in that requirements there apply to a wide range of fusion facilities, this volume is concerned mainly with 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 volume is oriented toward regulation in the Department of Energy (DOE) environment

  16. The construction of solid waste form test and inspection facility

    International Nuclear Information System (INIS)

    Park, Hun Hwee; Lee, Kang Moo; Jung, In Ha; Kim, Sung Hwan; Yoo, Jeong Woo; Lee, Jong Youl; Bae, Sang Min

    1988-01-01

    The solid waste form test and inspection facility is a facility to test and inspect the characteristics of waste forms, such as homogenity, mechanical structure, thermal behaviour, water resistance and leachability. Such kinds of characteristics in waste forms are required to meet a certain conditions for long-term storage or for final disposal of wastes. The facility will be used to evaluate safety for the disposal of wastes by test and inspection. At this moment, the efforts to search the most effective management of the radioactive wastes generated from power plants and radioisotope user are being executed by the people related to this field. Therefore, the facility becomes more significant tool because of its guidance of sucessfully converting wastes into forms to give a credit to the safety of waste disposal for managing the radioactive wastes. In addition the overall technical standards for inspecting of waste forms such as the standardized equipment and processes in the facility will be estabilished in the begining of 1990's when the project of waste management will be on the stream. Some of the items of the project have been standardized for the purpose of localization. In future, this facility will be utilized not only for the inspection of waste forms but also for the periodic decontamination apparatus by remote operation techniques. (Author)

  17. Radiological Operational Safety Verification for LILW Disposal Facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ju Youl [FNC Technology, SNU, Seoul (Korea, Republic of); Jeong, Seung Young; Kim, Byung Soo [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2011-10-15

    The successful implementation of radioactive waste repository program depends on scientific and technical aspects of excellent safety strategy as well as on societal aspects such as stakeholder acceptance and confidence. Monitoring is considered as key element in serving both ends. It covers all stages of the disposal process from site selection to institutional monitoring after the repository is closed. Basically, the purpose of the monitoring of radioactive waste disposal facility is not to reveal any increase of radioactivity due to the repository, but to provide reassurance and confirmation that the repository is fulfilling its passive safety purpose as an initial disposal concept and that long-term safety driven by regulatory requirements is ensured throughout the entire lifetime of disposal facility including post-closure phase. Five principal objectives of monitoring of geological disposal are summarized by IAEA-TECDOC-1208 as follows 1) Supporting management decisions in a staged programme of repository development: 2) Strengthening understanding of system behavior: 3) Societal decision making: 4) Accumulating an environmental database: 5) Nuclear safeguards (if repository contains fissile material, i.e., spent fuel or plutonium-rich waste) Based on the results of detailed studies of the above objectives and related phenomena, 6 categories of potential monitoring parameters are determined as follows: (1) degradation of repository structures, (2) behavior of the waste package and its associated buffer material, (3) near field chemical interactions between introduced materials, groundwater and host rock, (4) chemical and physical changes to the surrounding geosphere, (5) provision of an environmental database, and (6) nuclear safeguards. Typical monitoring parameters include temperature (heat), water level, pore-water and moisture content (groundwater), rock pressure, fractures, displacement and deformation (stress), water quality chemistry and dissolved

  18. Improvement of Safety Assessment Methodologies for Near Surface Disposal Facilities

    International Nuclear Information System (INIS)

    Batandjieva, B.; Torres-Vidal, C.

    2002-01-01

    The International Atomic Energy Agency (IAEA) Coordinated research program ''Improvement of Safety Assessment Methodologies for Near Surface Disposal Facilities'' (ISAM) has developed improved safety assessment methodology for near surface disposal facilities. The program has been underway for three years and has included around 75 active participants from 40 countries. It has also provided examples for application to three safety cases--vault, Radon type and borehole radioactive waste disposal facilities. The program has served as an excellent forum for exchange of information and good practices on safety assessment approaches and methodologies used worldwide. It also provided an opportunity for reaching broad consensus on the safety assessment methodologies to be applied to near surface low and intermediate level waste repositories. The methodology has found widespread acceptance and the need for its application on real waste disposal facilities has been clearly identified. The ISAM was finalized by the end of 2000, working material documents are available and an IAEA report will be published in 2002 summarizing the work performed during the three years of the program. The outcome of the ISAM program provides a sound basis for moving forward to a new IAEA program, which will focus on practical application of the safety assessment methodologies to different purposes, such as licensing radioactive waste repositories, development of design concepts, upgrading existing facilities, reassessment of operating repositories, etc. The new program will also provide an opportunity for development of guidance on application of the methodology that will be of assistance to both safety assessors and regulators

  19. Nuclear criticality safety program at the Fuel Cycle Facility

    International Nuclear Information System (INIS)

    Lell, R.M.; Fujita, E.K.; Tracy, D.B.; Klann, R.T.; Imel, G.R.; Benedict, R.W.; Rigg, R.H.

    1994-01-01

    The Fuel Cycle Facility (FCF) is designed to demonstrate the feasibility of a novel commercial-scale remote pyrometallurgical process for metallic fuels from liquid metal-cooled reactors and to show closure of the Integral Fast Reactor (IFR) fuel cycle. Requirements for nuclear criticality safety impose the most restrictive of the various constraints on the operation of FCF. The upper limits on batch sizes and other important process parameters are determined principally by criticality safety considerations. To maintain an efficient operation within appropriate safety limits, it is necessary to formulate a nuclear criticality safety program that integrates equipment design, process development, process modeling, conduct of operations, a measurement program, adequate material control procedures, and nuclear criticality analysis. The nuclear criticality safety program for FCF reflects this integration, ensuring that the facility can be operated efficiently without compromising safety. The experience gained from the conduct of this program in the Fuel cycle Facility will be used to design and safely operate IFR facilities on a commercial scale. The key features of the nuclear criticality safety program are described. The relationship of these features to normal facility operation is also described

  20. Criticality safety considerations. Integral Monitored Retrievable Storage (MRS) Facility

    International Nuclear Information System (INIS)

    1986-09-01

    This report summarizes the criticality analysis performed to address criticality safety concerns and to support facility design during the conceptual design phase of the Monitored Retrievable Storage (MRS) Facility. The report addresses the criticality safety concerns, the design features of the facility relative to criticality, and the results of the analysis of both normal operating and hypothetical off-normal conditions. Key references are provided (Appendix C) if additional information is desired by the reader. The MRS Facility design was developed and the related analysis was performed in accordance with the MRS Facility Functional Design Criteria and the Basis for Design. The detailed description and calculations are documented in the Integral MRS Facility Conceptual Design Report. In addition to the summary portion of this report, explanatary notes for various terms, calculation methodology, and design parameters are presented in Appendix A. Appendix B provides a brief glossary of technical terms

  1. 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).

  2. Safety of fuel cycle facilities. Topical issues paper no. 3

    International Nuclear Information System (INIS)

    Ranguelova, V.; Niehaus, F.; Delattre, D.

    2001-01-01

    A wide range of nuclear fuel cycle facilities are in operation. These installations process, use, store and dispose of radioactive material and cover: mining and milling, conversion, enrichment, fuel fabrication (including mixed oxide fuel), reactor, interim spent fuel storage, reprocessing, waste treatment and waste disposal facilities. For the purposes of this paper, reactors and waste disposal facilities are not considered. The term 'fuel cycle facilities' covers only the remainder of the installations listed above. The IAEA Secretariat maintains a database of fuel cycle facilities in its Member States. Known as the Nuclear Fuel Cycle Information System (NFCIS), it is available as an on-line service through the Internet. More than 500 such facilities have been reported under this system. The facilities are listed by facility type and operating status. Approximately one third of all of the facilities are located in developing States. About half of all facilities are reported to be operating, of which approximately 40% are operating in developing States. In addition, some 60 facilities are either in the design stage or under construction. Although the radioactive source term for most fuel cycle facilities is lower than the source term for reactors, which results in less severe consequences to the public from potential accidents at these fuel cycle installations, recent events at some fuel cycle facilities have given rise to public concern which has to be addressed adequately by national regulatory bodies and at the international level. Worldwide, operational experience feedback warrants improvements in the safety of these facilities. Some of the hazards are similar for reactor and non-reactor facilities. However, the differences between these installations give rise to specific safety concerns at fuel cycle facilities. In particular, these concerns include: criticality, radiation protection of workers, chemical hazards, fire and explosion hazards. It is recognized

  3. CP-50 calibration facility radiological safety assessment document

    International Nuclear Information System (INIS)

    Chilton, M.W.; Hill, R.L.; Eubank, B.F.

    1980-03-01

    The CP-50 Calibration Facility Radiological Safety Assessment document, prepared at the request of the Nevada Operations Office of the US Department of Energy to satisfy provisions of ERDA Manual Chapter 0531, presents design features, systems controls, and procedures used in the operation of the calibration facility. Site and facility characteristics and routine and non-routine operations, including hypothetical incidents or accidents are discussed and design factors, source control systems, and radiation monitoring considerations are described

  4. Test facilities for HTR, (2)

    International Nuclear Information System (INIS)

    Ishizuka, Hiroshi; Hayakawa, Hitoshi; Miki, Toshiya.

    1981-01-01

    The core of the multi-purpose high temperature gas-cooled experimental reactor is a circular column as a whole, in which the columns of stacked graphite blocks of hexagonal prism are arranged. The blocks in a column are doweled so as not to move horizontally, but adjacent columns vibrate while colliding mutually at the time of an earthquake because there is a gap between them. For the purpose to know the vibrating characteristics of a column surrounded by gap, Fuji Electric Co., Ltd., carried out the experiment. The tested column, the testing setup and the test result are reported. The distribution of flow rate in the core must be clarified, and the design data must be established early for confirming the feasibility of core design. The core structure tester was installed in Japan Atomic Energy Research Institute. The 1/2.75 scale model of the reactor bed was used, and the sealing performance of the block assemblies was tested. The sealing tester is related also to the distribution of flow rate in the core, and the basic performance of seal elements and the cross flow in fuel blocks were tested. The one-column tester and the seal element/two-column tester, the piping unit and the blower filter unit compose this tester. (Kako, I.)

  5. 400 Area/Fast Flux Test Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The 400 Area at Hanford is home primarily to the Fast Flux Test Facility (FFTF), a DOE-owned, formerly operating, 400-megawatt (thermal) liquid-metal (sodium)-cooled...

  6. Environmental Restoration Disposal Facility (Project W-296) Safety Assessment

    International Nuclear Information System (INIS)

    Armstrong, D.L.

    1994-08-01

    This Safety Assessment is based on information derived from the Conceptual Design Report for the Environmental Restoration Disposal Facility (DOE/RL 1994) and ancillary documentation developed during the conceptual design phase of Project W-296. The Safety Assessment has been prepared to support the Solid Waste Burial Ground Interim Safety Basis document. The purpose of the Safety Assessment is to provide an evaluation of the design to determine if the process, as proposed, will comply with US Department of Energy (DOE) Limits for radioactive and hazardous material exposures and be acceptable from an overall health and safety standpoint. The evaluation considered affects on the worker, onsite personnel, the public, and the environment

  7. Environmental Restoration Disposal Facility (Project W-296) Safety Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Armstrong, D.L.

    1994-08-01

    This Safety Assessment is based on information derived from the Conceptual Design Report for the Environmental Restoration Disposal Facility (DOE/RL 1994) and ancillary documentation developed during the conceptual design phase of Project W-296. The Safety Assessment has been prepared to support the Solid Waste Burial Ground Interim Safety Basis document. The purpose of the Safety Assessment is to provide an evaluation of the design to determine if the process, as proposed, will comply with US Department of Energy (DOE) Limits for radioactive and hazardous material exposures and be acceptable from an overall health and safety standpoint. The evaluation considered affects on the worker, onsite personnel, the public, and the environment.

  8. Construction of the two-phase critical flow test facility

    International Nuclear Information System (INIS)

    Chung, C. H.; Chang, S. K.; Park, H. S.; Min, K. H.; Choi, N. H.; Kim, C. H.; Lee, S. H.; Kim, H. C.; Chang, M. H.

    2002-03-01

    The two-phase critical test loop facility has been constructed in the KAERI engineering laboratory for the simulation of small break loss of coolant accident entrained with non-condensible gas of SMART. The test facility can operate at 12 MPa of pressure and 0 to 60 C of sub-cooling with 0.5 kg/s of non- condensible gas injection into break flow, and simulate up to 20 mm of pipe break. Main components of the test facility were arranged such that the pressure vessel containing coolant, a test section simulating break and a suppression tank inter-connected with pipings were installed vertically. As quick opening valve opens, high pressure/temperature coolant flows through the test section forming critical two-phase flow into the suppression tank. The pressure vessel was connected to two high pressure N2 gas tanks through a control valve to control pressure in the pressure vessel. Another N2 gas tank was also connected to the test section for the non-condensible gas injection. The test facility operation was performed on computers supported with PLC systems installed in the control room, and test data such as temperature, break flow rate, pressure drop across test section, gas injection flow rate were all together gathered in the data acquisition system for further data analysis. This test facility was classified as a safety related high pressure gas facility in law. Thus the loop design documentation was reviewed, and inspected during construction of the test loop by the regulatory body. And the regulatory body issued permission for the operation of the test facility

  9. Technical Safety Requirements for the Waste Storage Facilities May 2014

    Energy Technology Data Exchange (ETDEWEB)

    Laycak, D. T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2014-04-16

    This document contains the Technical Safety Requirements (TSR) for the Radioactive and Hazardous Waste Management (RHWM) WASTE STORAGE FACILITIES, which include Area 625 (A625) and the Building 693 (B693) Yard Area of the Decontamination and Waste Treatment Facility (DWTF) at LLNL. The TSRs constitute requirements for safe operation of the WASTE STORAGE FACILITIES. These TSRs are derived from the Documented Safety Analyses for the Waste Storage Facilities (DSA) (LLNL 2011). The analysis presented therein concluded that the WASTE STORAGE FACILITIES are low-chemical hazard, Hazard Category 2 non-reactor nuclear facilities. The TSRs consist primarily of inventory limits and controls to preserve the underlying assumptions in the hazard and accident analyses. Further, appropriate commitments to safety programs are presented in the administrative controls sections of the TSRs. The WASTE STORAGE FACILITIES are used by RHWM to handle and store hazardous waste, TRANSURANIC (TRU) WASTE, LOW-LEVEL WASTE (LLW), mixed waste, California combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL as well as small amounts of waste from other DOE facilities, as described in the DSA. In addition, several minor treatments (e.g., size reduction and decontamination) are carried out in these facilities.

  10. Technical Safety Requirements for the Waste Storage Facilities May 2014

    International Nuclear Information System (INIS)

    Laycak, D. T.

    2014-01-01

    This document contains the Technical Safety Requirements (TSR) for the Radioactive and Hazardous Waste Management (RHWM) WASTE STORAGE FACILITIES, which include Area 625 (A625) and the Building 693 (B693) Yard Area of the Decontamination and Waste Treatment Facility (DWTF) at LLNL. The TSRs constitute requirements for safe operation of the WASTE STORAGE FACILITIES. These TSRs are derived from the Documented Safety Analyses for the Waste Storage Facilities (DSA) (LLNL 2011). The analysis presented therein concluded that the WASTE STORAGE FACILITIES are low-chemical hazard, Hazard Category 2 non-reactor nuclear facilities. The TSRs consist primarily of inventory limits and controls to preserve the underlying assumptions in the hazard and accident analyses. Further, appropriate commitments to safety programs are presented in the administrative controls sections of the TSRs. The WASTE STORAGE FACILITIES are used by RHWM to handle and store hazardous waste, TRANSURANIC (TRU) WASTE, LOW-LEVEL WASTE (LLW), mixed waste, California combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL as well as small amounts of waste from other DOE facilities, as described in the DSA. In addition, several minor treatments (e.g., size reduction and decontamination) are carried out in these facilities.

  11. 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

  12. 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

  13. 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.)

  14. Waste Encapsulation and Storage Facility interim operational safety requirements

    CERN Document Server

    Covey, L I

    2000-01-01

    The Interim Operational Safety Requirements (IOSRs) for the Waste Encapsulation and Storage Facility (WESF) define acceptable conditions, safe boundaries, bases thereof, and management or administrative controls required to ensure safe operation during receipt and inspection of cesium and strontium capsules from private irradiators; decontamination of the capsules and equipment; surveillance of the stored capsules; and maintenance activities. Controls required for public safety, significant defense-in-depth, significant worker safety, and for maintaining radiological consequences below risk evaluation guidelines (EGs) are included.

  15. Sultan - forced flow, high field test facility

    International Nuclear Information System (INIS)

    Horvath, I.; Vecsey, G.; Weymuth, P.; Zellweger, J.

    1981-01-01

    Three European laboratories: CNEN (Frascati, I) ECN (Petten, NL) and SIN (Villigen, CH) decided to coordinate their development efforts and to install a common high field forced flow test facility at Villigen Switzerland. The test facility SULTAN (Supraleiter Testanlage) is presently under construction. As a first step, an 8T/1m bore solenoid with cryogenic periphery will be ready in 1981. The cryogenic system, data acquisition system and power supplies which are contributed by SIN are described. Experimental feasibilities, including cooling, and instrumentation are reviewed. Progress of components and facility construction is described. Planned extension of the background field up to 12T by insert coils is outlined. 5 refs

  16. Advanced Control Test Operation (ACTO) facility

    International Nuclear Information System (INIS)

    Ball, S.J.

    1987-01-01

    The Advanced Control Test Operation (ACTO) project, sponsored by the US Department of Energy (DOE), is being developed to enable the latest modern technology, automation, and advanced control methods to be incorporated into nuclear power plants. The facility is proposed as a national multi-user center for advanced control development and testing to be completed in 1991. The facility will support a wide variety of reactor concepts, and will be used by researchers from Oak Ridge National Laboratory (ORNL), plus scientists and engineers from industry, other national laboratories, universities, and utilities. ACTO will also include telecommunication facilities for remote users

  17. Regulatory role and approach of BARC Safety Council in safety and occupational health in BARC facilities

    International Nuclear Information System (INIS)

    Rajdeep; Jayarajan, K.; Taly, Y.K.

    2016-01-01

    Bhabha Atomic Research Centre is involved in multidisciplinary research and developmental activities, related to peaceful use of nuclear energy and its societal benefits. In order to achieve high level of performance of these facilities, the best efforts are made to maintain good health of the plant personnel and good working conditions. BARC Safety Council (BSC), which is the regulatory body for BARC facilities, regulates radiation safety, industrial safety and surveillance of occupational health, by implementing various rules and guidelines in BARC facilities. BARC Safety framework consists of various committees in a 3-tier system. The first tier is BSC, which is the apex body authorized for issuing directives, permissions, consents and authorizations. It is having responsibility of ensuring protection and safety of public, environment, personnel and facilities of BARC through enforcement of radiation protection and industrial safety programmes. Besides the 18 committees in 2"n"d tier, there are 6 other expert committees which assist in functioning of BSC. (author)

  18. Preliminary safety analysis report for the Waste Characterization Facility

    International Nuclear Information System (INIS)

    1994-10-01

    This safety analysis report outlines the safety concerns associated with the Waste Characterization Facility located in the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. The three main objectives of the report are to: define and document a safety basis for the Waste Characterization Facility activities; demonstrate how the activities will be carried out to adequately protect the workers, public, and environment; and provide a basis for review and acceptance of the identified risk that the managers, operators, and owners will assume. 142 refs., 38 figs., 39 tabs

  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. Critical experiments facility and criticality safety programs at JAERI

    International Nuclear Information System (INIS)

    Kobayashi, Iwao; Tachimori, Shoichi; Takeshita, Isao; Suzaki, Takenori; Miyoshi, Yoshinori; Nomura, Yasushi

    1985-10-01

    The nuclear criticality safety is becoming a key point in Japan in the safety considerations for nuclear installations outside reactors such as spent fuel reprocessing facilities, plutonium fuel fabrication facilities, large scale hot alboratories, and so on. Especially a large scale spent fuel reprocessing facility is being designed and would be constructed in near future, therefore extensive experimental studies are needed for compilation of our own technical standards and also for verification of safety in a potential criticality accident to obtain public acceptance. Japan Atomic Energy Research Institute is proceeding a construction program of a new criticality safety experimental facility where criticality data can be obtained for such solution fuels as mainly handled in a reprocessing facility and also chemical process experiments can be performed to investigate abnormal phenomena, e.g. plutonium behavior in solvent extraction process by using pulsed colums. In FY 1985 detail design of the facility will be completed and licensing review by the government would start in FY 1986. Experiments would start in FY 1990. Research subjects and main specifications of the facility are described. (author)

  1. Massachusetts Large Blade Test Facility Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Rahul Yarala; Rob Priore

    2011-09-02

    Project Objective: The Massachusetts Clean Energy Center (CEC) will design, construct, and ultimately have responsibility for the operation of the Large Wind Turbine Blade Test Facility, which is an advanced blade testing facility capable of testing wind turbine blades up to at least 90 meters in length on three test stands. Background: Wind turbine blade testing is required to meet international design standards, and is a critical factor in maintaining high levels of reliability and mitigating the technical and financial risk of deploying massproduced wind turbine models. Testing is also needed to identify specific blade design issues that may contribute to reduced wind turbine reliability and performance. Testing is also required to optimize aerodynamics, structural performance, encourage new technologies and materials development making wind even more competitive. The objective of this project is to accelerate the design and construction of a large wind blade testing facility capable of testing blades with minimum queue times at a reasonable cost. This testing facility will encourage and provide the opportunity for the U.S wind industry to conduct more rigorous testing of blades to improve wind turbine reliability.

  2. Repetitively pulsed material testing facility

    International Nuclear Information System (INIS)

    Zucker, O.; Bostick, W.; Gullickson, R; Long, J.; Luce, J.; Sahlin, H.

    1975-01-01

    A continuously operated, 1 pps, dense-plasma-focus device capable of delivering a minimum of 10 15 neutrons per pulse for material testing purposes is described. Moderate scaling from existing results is sufficient to provide 2 x 10 13 n/cm 2 .s to a suitable target. The average power consumption, which has become a major issue as a result of the energy crisis, is analyzed with respect to other plasma devices and is shown to be highly favorable. A novel approach to the capacitor bank and switch design allowing repetitive operation is discussed. (U.S.)

  3. Detector development and test facility

    International Nuclear Information System (INIS)

    Reeder, D.D.

    1993-01-01

    Following the ideas presented in the proposal to the DoE, we have begun to acquire the equipment needed to design, develop construct and test the electronic and mechanical features of detectors used in High Energy Physics Experiments. A guiding principle for the effort is to achieve integrated electronic and mechanical designs which meet the demanding specifications of the modern hadron collider environment yet minimize costs. This requires state of the art simulation of signal processing as well as detailed calculations of heat transfer and finite element analysis of structural integrity

  4. Repetitively pulsed material testing facility

    International Nuclear Information System (INIS)

    Zucker, O.; Bostick, W.; Gullickson, R.; Long, J.; Luce, J.; Sahlin, H.

    1975-01-01

    A continuously operated, 1 pps, dense-plasma-focus device capable of delivering a minimum of 10 15 neutrons per pulse for material testing purposes is described. Moderate scaling from existing results is sufficient to provide 2 x 10 13 n/cm 2 . s to a suitable target. The average power consumption, which has become a major issue as a result of the energy crisis, is analyzed with respect to other plasma devices and is shown to be highly favorable. A novel approach to the capacitor bank and switch design allowing repetitive operation is discussed

  5. Fusion Materials Irradiation Test Facility: experimental capabilities and test matrix

    International Nuclear Information System (INIS)

    Opperman, E.K.

    1982-01-01

    This report describes the experimental capabilities of the Fusion Materials Irradiation Test Facility (FMIT) and reference material specimen test matrices. The description of the experimental capabilities and the test matrices has been updated to match the current single test cell facility ad assessed experimenter needs. Sufficient detail has been provided so that the user can plan irradiation experiments and conceptual hardware. The types of experiments, irradiation environment and support services that will be available in FMIT are discussed

  6. Directory of transport packaging test facilities

    International Nuclear Information System (INIS)

    1983-08-01

    Radioactive materials are transported in packagings or containers which have to withstand certain tests depending on whether they are Type A or Type B packagings. In answer to a request by the International Atomic Energy Agency, 13 Member States have provided information on the test facilities and services existing in their country which can be made available for use by other states by arrangement for testing different kinds of packagings. The directory gives the technical information on the facilities, the services, the tests that can be done and in some cases even the financial arrangement is included

  7. Business of Nuclear Safety Analysis Office, Nuclear Technology Test Center

    International Nuclear Information System (INIS)

    Hayakawa, Masahiko

    1981-01-01

    The Nuclear Technology Test Center established the Nuclear Safety Analysis Office to execute newly the works concerning nuclear safety analysis in addition to the works related to the proving tests of nuclear machinery and equipments. The regulations for the Nuclear Safety Analysis Office concerning its organization, business and others were specially decided, and it started the business formally in August, 1980. It is a most important subject to secure the safety of nuclear facilities in nuclear fuel cycle as the premise of developing atomic energy. In Japan, the strict regulation of safety is executed by the government at each stage of the installation, construction, operation and maintenance of nuclear facilities, based on the responsibility for the security of installers themselves. The Nuclear Safety Analysis Office was established as the special organ to help the safety examination related to the installation of nuclear power stations and others by the government. It improves and puts in order the safety analysis codes required for the cross checking in the safety examination, and carries out safety analysis calculation. It is operated by the cooperation of the Science and Technology Agency and the Agency of Natural Resources and Energy. The purpose of establishment, the operation and the business of the Nuclear Safety Analysis Office, the plan of improving and putting in order of analysis codes, and the state of the similar organs in foreign countries are described. (Kako, I.)

  8. Automation of electromagnetic compatability (EMC) test facilities

    Science.gov (United States)

    Harrison, C. A.

    1986-01-01

    Efforts to automate electromagnetic compatibility (EMC) test facilities at Marshall Space Flight Center are discussed. The present facility is used to accomplish a battery of nine standard tests (with limited variations) deigned to certify EMC of Shuttle payload equipment. Prior to this project, some EMC tests were partially automated, but others were performed manually. Software was developed to integrate all testing by means of a desk-top computer-controller. Near real-time data reduction and onboard graphics capabilities permit immediate assessment of test results. Provisions for disk storage of test data permit computer production of the test engineer's certification report. Software flexibility permits variation in the tests procedure, the ability to examine more closely those frequency bands which indicate compatibility problems, and the capability to incorporate additional test procedures.

  9. Ice condenser testing facility and plans

    International Nuclear Information System (INIS)

    Kannberg, L.D.; Ross, B.A.; Eschbach, E.J.; Ligotke, M.W.

    1987-01-01

    A facility is being constructed to experimentally validate the ICEDF computer code. The code was developed to estimate the extent of fission product retention in the ice compartments of pressurized water reactor ice condenser containment systems during severe accidents. The design and construction of the facility is based on a test design that addresses the validation needs of the code for conditions typical of those expected to occur during severe pressurized water reactor accidents. Detailed facility design has followed completion of a test design (i.e., assembled test cases each involving a different set of aerosol and thermohydraulic flow conditions). The test design was developed with the aid of statistical test design software and was scrutinized for applicability with the aid of ICEDF simulations. The test facility will incorporate a small section of a prototypic ice condenser (e.g., a cross section comprising the equivalent of four 1-ft-diameter ice baskets to their full prototypic height of 48 ft). The development of the test design, the detailed facility design, and the construction progress are described in this paper

  10. Fire Safety. Managing School Facilities, Guide 6.

    Science.gov (United States)

    Department for Education and Employment, London (England). Architects and Building Branch.

    This booklet discusses how United Kingdom schools can manage fire safety and minimize the risk of fire. The guide examines what legislation school buildings must comply with and covers the major risks. It also describes training and evacuation procedures and provides guidance on fire precautions, alarm systems, fire fighting equipment, and escape…

  11. 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.

  12. 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

  13. Technical Safety Requirements for the Gamma Irradiation Facility (GIF)

    CERN Document Server

    Mahn, J A E M J G

    2003-01-01

    This document provides the Technical Safety Requirements (TSR) for the Sandia National Laboratories Gamma Irradiation Facility (GIF). The TSR is a compilation of requirements that define the conditions, the safe boundaries, and the administrative controls necessary to ensure the safe operation of a nuclear facility and to reduce the potential risk to the public and facility workers from uncontrolled releases of radioactive or other hazardous materials. These requirements constitute an agreement between DOE and Sandia National Laboratories management regarding the safe operation of the Gamma Irradiation Facility.

  14. Evolution of nuclear safety regulation for BARC Facilities

    International Nuclear Information System (INIS)

    Jayarajan, K.; Taly, Y.K.

    2017-01-01

    Safety programmes in BARC stared during the formative years and grown its stature, as the years passed by. Seventeen years of BSC, with one hundred meetings, have been quite eventful with several achievements. BSC could bring all facilities of BARC under its safety umbrella and could streamline many safety and regulatory activities. BSC aims at incident free operation of all facilities and protection of the workers, the public, the environment from radiation and other hazards. Although, incidents could not be entirely prevented, BSC have taken every event as a lesson and used the experience for improving safety. Safety enhancement is an endless journey, which has to be performed by joining hands of the managers, designers, manufacturers, inspectors and operators, in addition to the regulators

  15. Recommissioning the K-1600 seismic test facility

    International Nuclear Information System (INIS)

    Wynn, C.C.; Brewer, D.W.

    1991-01-01

    The Center for Natural Phenomena Engineering (CNPE) was established under the technical direction of Dr. James E. Beavers with a mandate to assess, by analyses and testing, the seismic capacity of building structures that house sensitive processes at the Oak Ridge Y-12 Plant. This mandate resulted in a need to recommission the K-1600 Seismic Test Facility (STF) at the Oak Ridge K-25 Site, which had been shutdown for 6 years. This paper documents the history of the facility and gives some salient construction, operation, and performance details of its 8-ton, 20-foot center of gravity payload biaxial seismic simulator. A log of activities involved in the restart of this valuable resource is included as Table 1. Some of the problems and solutions associated with recommissioning the facility under a relatively limited budget are included. The unique attributes of the shake table are discussed. The original mission and performance requirements are compared to current expanded mission and performance capabilities. Potential upgrades to further improve the capabilities of the test facility as an adjunct to the CNPE are considered. Additional uses for the facility are proposed, including seismic qualification testing of devices unique to enrichment technologies and associated hazardous waste treatment and disposal processes. In summary, the STF restart in conjunction with CNPE has added a vital, and unique facility to the list of current national resources utilized for earthquake engineering research and development

  16. Construction of solid waste form test facility

    International Nuclear Information System (INIS)

    Park, Hyun Whee; Lee, Kang Moo; Koo, Jun Mo; Jung, In Ha; Lee, Jong Ryeul; Kim, Sung Whan; Bae, Sang Min; Cho, Kang Whon; Sung, Suk Jong

    1989-02-01

    The Solid Waste Form Test Facility (SWFTF) is now construction at DAEDUCK in Korea. In SWFTF, the characteristics of solidified waste products as radiological homogeneity, mechanical and thermal property, water resistance and lechability will be tested and evaluated to meet conditions for long-term storage or final disposal of wastes. The construction of solid waste form test facility has been started with finishing its design of a building and equipments in Sep. 1984, and now building construction is completed. Radioactive gas treatment system, extinguishers, cooling and heating system for the facility, electrical equipments, Master/Slave manipulator, power manipulator, lead glass and C.C.T.V. has also been installed. SWFTF will be established in the beginning of 1990's. At this report, radiation shielding door, nondestructive test of the wall, instrumentation system for the utility supply system and cell lighting system are described. (Author)

  17. Design of a hydrogen test facility

    International Nuclear Information System (INIS)

    Morgan, M.J.; Beam, J.E.; Sehmbey, M.S.; Pais, M.R.; Chow, L.C.; Hahn, O.J.

    1992-01-01

    The Air Force has sponsored a program at the University of Kentucky which will lead to a better understanding of the thermal and fluid instabilities during blowdown of supercritical fluids at cryogenic temperatures. An integral part of that program is the design and construction of that hydrogen test facility. This facility will be capable of providing supercritical hydrogen at 30 bars and 35 K at a maximum flow rate of 0.1 kg/s for 90 seconds. Also presented here is an extension of this facility to accommodate the use of supercritical helium

  18. Status of superconducting RF test facility (STF)

    International Nuclear Information System (INIS)

    Hayano, Hitoshi

    2005-01-01

    A superconducting technology was recommended for the main linac design of the International Linear Collider (ILC) by the International Technology Recommendation Panel (ITRP). The basis for this design has been developed and tested at DESY, and R and D is progressing at many laboratories around the world including DESY, Orsay, KEK, FNAL, SLAC, Cornell, and JLAB. In order to promote Asian SC-technology for ILC, construction of a test facility in KEK was discussed and decided. The role and status of the superconducting RF test facility (STF) is reported in this paper. (author)

  19. Criticality safety research on nuclear fuel cycle facility

    Energy Technology Data Exchange (ETDEWEB)

    Miyoshi, Yoshinori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2004-07-01

    This paper present d s current status and future program of the criticality safety research on nuclear fuel cycle made by Japan Atomic Energy Research Institute. Experimental research on solution fuel treated in reprocessing plant has been performed using two critical facilities, STACY and TRACY. Fundamental data of static and transient characteristics are accumulated for validation of criticality safety codes. Subcritical measurements are also made for developing a monitoring system for criticality safety. Criticality safety codes system for solution and power system, and evaluation method related to burnup credit are developed. (author)

  20. A negative ion source test facility

    Energy Technology Data Exchange (ETDEWEB)

    Melanson, S.; Dehnel, M., E-mail: morgan@d-pace.com; Potkins, D.; Theroux, J.; Hollinger, C.; Martin, J.; Stewart, T.; Jackle, P.; Withington, S. [D-Pace, Inc., P.O. Box 201, Nelson, British Columbia V1L 5P9 (Canada); Philpott, C.; Williams, P.; Brown, S.; Jones, T.; Coad, B. [Buckley Systems Ltd., 6 Bowden Road, Mount Wellington, Auckland 1060 (New Zealand)

    2016-02-15

    Progress is being made in the development of an Ion Source Test Facility (ISTF) by D-Pace Inc. in collaboration with Buckley Systems Ltd. in Auckland, NZ. The first phase of the ISTF is to be commissioned in October 2015 with the second phase being commissioned in March 2016. The facility will primarily be used for the development and the commercialization of ion sources. It will also be used to characterize and further develop various D-Pace Inc. beam diagnostic devices.

  1. Manual for operation of the multipurpose thermalhydraulic test facility TOPFLOW (Transient Two Phase Flow Test Facility)

    International Nuclear Information System (INIS)

    Beyer, M.; Carl, H.; Schuetz, H.; Pietruske, H.; Lenk, S.

    2004-07-01

    The Forschungszentrum Rossendorf (FZR) e. V. is constructing a new large-scale test facility, TOPFLOW, for thermalhydraulic single effect tests. The acronym stands for transient two phase flow test facility. It will mainly be used for the investigation of generic and applied steady state and transient two phase flow phenomena and the development and validation of models of computational fluid dynamic (CFD) codes. The manual of the test facility must always be available for the staff in the control room and is restricted condition during operation of personnel and also reconstruction of the facility. (orig./GL)

  2. Startup of large coil test facility

    International Nuclear Information System (INIS)

    Haubenreich, P.N.; Bohanan, R.E.; Fietz, W.A.; Luton, J.N.; May, J.R.

    1984-01-01

    The Large Coil Test Facility (LCTF) is being used to test superconducting toroidal field coils about one-third the size of those for INTOR. Data were obtained on performance of refrigerator, helium distribution, power supplies, controls, and data acquisition systems and on the acoustic emission, voltages, currents, and mechanical strains during charging and discharging the coils. (author)

  3. SULTAN test facility: Summary of recent results

    International Nuclear Information System (INIS)

    Stepanov, Boris; Bruzzone, Pierluigi; Sedlak, Kamil; Croari, Giancarlo

    2013-01-01

    The test campaigns of the ITER conductors in the SULTAN test facility re-started in December 2011 after three months break. The main focus of the activities is about the qualification tests of the Central Solenoid (CS) conductors, with three different samples for a total six variations of strand suppliers and cable layouts. In 2012, five Toroidal Field (TF) conductor samples have also been tested as part of the supplier and process qualification phase of the European, Korean, Chinese and Russian Federation Agencies. A summary of the test results for all the ITER samples tested in the last period is presented, including an updated statistics of the broad transition, the performance degradation and the impact of layout variations. The role of SULTAN test facility during the ITER construction is reviewed, and the load of work for the next three years is anticipated

  4. Radiation safety of gamma and electron irradiation facilities

    International Nuclear Information System (INIS)

    1992-01-01

    There are currently some 160 gamma irradiation facilities and over 600 electron beam facilities in operation throughout virtually all Member States of the IAEA. The most widespread uses of these facilities are for the sterilization of medical and pharmaceutical products, the preservation of foodstuffs, polymer synthesis and modification, and the eradication of insect infestation. The safety record of this industry has been very good. Nevertheless, there is a potential for accidents with serious consequences. Gamma and electron beam facilities produce very high dose rates during irradiation, so that a person accidentally present in the irradiation chamber can receive a lethal dose within minutes or seconds. Precautions against uncontrolled entry must therefore be taken. Furthermore, gamma irradiation facilities contain large amounts of radioactivity and if the mechanism for retracting the source is damaged, the source may remain exposed, inhibiting direct access to carry out remedial work. Contamination can result from corroded or damaged sources, and decontamination can be very expensive. These aspects clearly indicate the need to achieve a high degree of safety and reliability in the facilities. This can be accomplished by effective quality control together with careful design, manufacture, installation, operation and decommissioning. The guidance in this Safety Series publication is intended for competent authorities responsible for regulating the use of radiation sources as well as the manufacturers, suppliers, installers and users of gamma and electron beam facilities. 20 refs, 6 figs

  5. IAEA safety requirements for safety assessment of fuel cycle facilities and activities

    International Nuclear Information System (INIS)

    Jones, G.

    2013-01-01

    The IAEA's Statute authorises the Agency to establish standards of safety for protection of health and minimisation of danger to life and property. In that respect, the IAEA has established a Safety Fundamentals publication which contains ten safety principles for ensuring the protection of workers, the public and the environment from the harmful effects of ionising radiation. A number of these principles require safety assessments to be carried out as a means of evaluating compliance with safety requirements for all nuclear facilities and activities and to determine the measures that need to be taken to ensure safety. The safety assessments are required to be carried out and documented by the organisation responsible for operating the facility or conducting the activity, are to be independently verified and are to be submitted to the regulatory body as part of the licensing or authorisation process. In addition to the principles of the Safety Fundamentals, the IAEA establishes requirements that must be met to ensure the protection of people and the environment and which are governed by the principles in the Safety Fundamentals. The IAEA's Safety Requirements publication 'Safety Assessment for Facilities and Activities', establishes the safety requirements that need to be fulfilled in conducting and maintaining safety assessments for the lifetime of facilities and activities, with specific attention to defence in depth and the requirement for a graded approach to the application of these safety requirements across the wide range of fuel cycle facilities and activities. Requirements for independent verification of the safety assessment that needs to be carried out by the operating organisation, including the requirement for the safety assessment to be periodically reviewed and updated are also covered. For many fuel cycle facilities and activities, environmental impact assessments and non-radiological risk assessments will be required. The

  6. Research and test facilities required in nuclear science and technology

    International Nuclear Information System (INIS)

    2009-01-01

    Experimental facilities are essential research tools both for the development of nuclear science and technology and for testing systems and materials which are currently being used or will be used in the future. As a result of economic pressures and the closure of older facilities, there are concerns that the ability to undertake the research necessary to maintain and to develop nuclear science and technology may be in jeopardy. An NEA expert group with representation from ten member countries, the International Atomic Energy Agency and the European Commission has reviewed the status of those research and test facilities of interest to the NEA Nuclear Science Committee. They include facilities relating to nuclear data measurement, reactor development, neutron scattering, neutron radiography, accelerator-driven systems, transmutation, nuclear fuel, materials, safety, radiochemistry, partitioning and nuclear process heat for hydrogen production. This report contains the expert group's detailed assessment of the current status of these nuclear research facilities and makes recommendations on how future developments in the field can be secured through the provision of high-quality, modern facilities. It also describes the online database which has been established by the expert group which includes more than 700 facilities. (authors)

  7. The TOPFLOW multi-purpose thermohydraulic test facility

    International Nuclear Information System (INIS)

    Schaffrath, Andreas; Kruessenberg, A.-K.; Weiss, F.-P.; Prasser, H.-M.

    2002-01-01

    The TOPFLOW (Transient Two Phase Flow Test Facility) multi-purpose thermohydraulic test facility is being built for studies of steady-state and transient flow phenomena in two-phase flows, and for the development and validation of the models contained in CFD (Computational Fluid Dynamics) codes. The facility is under construction at the Institute for Safety Research of the Rossendorf Research Center (FZR). It will be operated together with the Dresden Technical University and the Zittau/Goerlitz School for Technology, Economics and Social Studies within the framework of the Nuclear Technology Competence Preservation Program. TOPFLOW, with its test sections and its flexible concept, is available as an attractive facility also to users from all European countries. Experiments are planned in these fields, among others: - Transient two-phase flows in vertical and horizontal pipes and pipes of any inclination as well as in geometries typical of nuclear reactors (annulus, hot leg). - Boiling in large vessels and water pools (measurements of steam generation, 3D steam content distribution, turbulence, temperature stratification). - Test of passive components and safety systems. - Condensation in horizontal pipes in the absence and presence of non-condensable gases. The construction phase of TOPFLOW has been completed more or less on schedule. Experiments can be started after a commissioning phase in the 3rd quarter of 2002. (orig.) [de

  8. Decommissioning of Facilities. General Safety Requirements. Pt. 6

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-07-15

    Decommissioning is the last step in the lifetime management of a facility. It must also be considered during the design, construction, commissioning and operation of facilities. This publication establishes requirements for the safe decommissioning of a broad range of facilities: nuclear power plants, research reactors, nuclear fuel cycle facilities, facilities for processing naturally occurring radioactive material, former military sites, and relevant medical, industrial and research facilities. It addresses all the aspects of decommissioning that are required to ensure safety, aspects such as roles and responsibilities, strategy and planning for decommissioning, conduct of decommissioning actions and termination of the authorization for decommissioning. It is intended for use by those involved in policy development, regulatory control and implementation of decommissioning.

  9. Decommissioning of Facilities. General Safety Requirements. Pt. 6 (Spanish Edition)

    International Nuclear Information System (INIS)

    2017-01-01

    Decommissioning is the last step in the lifetime management of a facility. It must also be considered during the design, construction, commissioning and operation of facilities. This publication establishes requirements for the safe decommissioning of a broad range of facilities: nuclear power plants, research reactors, nuclear fuel cycle facilities, facilities for processing naturally occurring radioactive material, former military sites, and relevant medical, industrial and research facilities. It addresses all the aspects of decommissioning that are required to ensure safety, aspects such as roles and responsibilities, strategy and planning for decommissioning, conduct of decommissioning actions and termination of the authorization for decommissioning. It is intended for use by those involved in policy development, regulatory control and implementation of decommissioning.

  10. Decommissioning of Facilities. General Safety Requirements. Pt. 6 (Russian Edition)

    International Nuclear Information System (INIS)

    2015-01-01

    Decommissioning is the last step in the lifetime management of a facility. It must also be considered during the design, construction, commissioning and operation of facilities. This publication establishes requirements for the safe decommissioning of a broad range of facilities: nuclear power plants, research reactors, nuclear fuel cycle facilities, facilities for processing naturally occurring radioactive material, former military sites, and relevant medical, industrial and research facilities. It addresses all the aspects of decommissioning that are required to ensure safety, aspects such as roles and responsibilities, strategy and planning for decommissioning, conduct of decommissioning actions and termination of the authorization for decommissioning. It is intended for use by those involved in policy development, regulatory control and implementation of decommissioning

  11. Data used for safety assessment of reprocessing facilities

    International Nuclear Information System (INIS)

    Nomura, Yasushi; Suzuki, Atsuyuki; Kanagawa, Akira

    1990-08-01

    For safety assessment of a reprocessing facility, it is important to know performance of radioactive materials in their accidental release and transfer. Accordingly, it is necessary to collect and prepare data for use in analyses for their performance. In JAERI, experiments such as for data acquisition, for source-term evaluation and for radioactive material transfer, are now planned to be performed. Prior to these experiments, it is decided to investigate data in use for accidental safety assessment of reprocessing plants and their based experimental data, thus to make it possible to recommend reasonable values for safety analysis parameters by evaluating the investigated results, to select the experimental items, to edit a safety assessment handbook and so on. In this line of objectives, JAERI rewarded a two-year contract of investigation to Nuclear Safety Research Association, to make a working group under a special committee on data investigation for reprocessing facility safety assessment. This report is a collection of results reviewed and checked by the working group. The contents consist of two parts, one for investigation and review of data used for safety assessment of domestic or oversea reprocessing facilities, and the other for investigation, review and evaluation of ANSI recommended American standard data reported by E. Walker together with their based experimental data resorting to the original referred reports. (author)

  12. Radiation safety and regulatory aspects in Medical Facilities

    International Nuclear Information System (INIS)

    Banerjee, Sharmila

    2017-01-01

    Radiation safety and regulatory aspect of medical facilities are relevant in the context where radiation is used in providing healthcare to human patients. These include facilities, which carry out radiological procedures in diagnostic radiology, including dentistry, image-guided interventional procedures, nuclear medicine, and radiation therapy. The safety regulations provide recommendations and guidance on meeting the requirements for the safe use of radiation in medicine. The different safety aspects which come under its purview are the personnel involved in medical facilities where radiological procedures are performed which include the medical practitioners, radiation technologists, medical physicists, radiopharmacists, radiation protection and over and above all the patients. Regulatory aspects cover the guidelines provided by ethics committees, which regulate the administration of radioactive formulation in human patients. Nuclear medicine is a modality that utilizes radiopharmaceuticals either for diagnosis of physiological disorders related to anatomy, physiology and patho-physiology and for diagnosis and treatment of cancer

  13. Safety analysis of the existing 850 Firing Facility

    International Nuclear Information System (INIS)

    Odell, B.N.

    1986-01-01

    A safety analysis was performed to determine if normal operations and/or potential accidents at the 850 Firing Facility at Site 300 could present undue hazards to the general public, personnel at Site 300, or have an adverse effect on the environment. The normal operations and credible accidents that might have an effect on these facilities or have off-site consequences were considered. It was determined by this analysis that all but one of the hazards were either low or of the type or magnitude routinely encountered and/or accepted by the public. The exception was explosives, which was classified as a moderate hazard per the requirements given in DOE Order 5481.1A. This safety analysis concluded that the operation at this facility will present no undue risk to the health and safety of LLNL employees or the public

  14. Safety analysis of the existing 851 Firing Facility

    International Nuclear Information System (INIS)

    Odell, B.N.

    1986-01-01

    A safety analysis was performed to determine if normal operations and/or potential accidents at the 851 Firing Facility at Site 300 could present undue hazards to the general public, personnel at Site 300, or have an adverse effect on the environment. The normal operations and credible accidents that might have an effect on these facilities or have off-site consequences were considered. It was determined by this analysis that all but two of the hazards were either low or of the type or magnitude routinely encountered and/or accepted by the public. The exceptions were the linear accelerator and explosives, which were classified as moderate hazards per the requirements given in DOE Order 5481.1A. This safety analysis concluded that the operation at this facility will present no undue risk to the health and safety of LLNL employees or the public

  15. 2-MW plasmajet facility thermal tests of concrete

    International Nuclear Information System (INIS)

    Goin, K.L.

    1977-07-01

    A test was made in the 2-Megawatt Plasmajet Facility to obtain experimental data relative to the thermal response of concrete to incident heat flux. 14.6 cm diameter by 8.0 cm long concrete cylinders were positioned in a supersonic flow of heated nitrogen from an arc heater. The end of the concrete cylinders impacted by the flow were subjected to heat fluxes in the range of 0.13 to 0.35 kW/cm 2 . Measurements included cold wall surface heat flux and pressure distributions, surface and indepth temperatures, ablation rates, and surface emission spectrographs. The test was part of the Sandia light water reactor safety research program and complements similar tests made in the Radiant Heat Facility at heat fluxes from 0.03 to 0.12 kW/cm 2 . A description of the tests and a tabulation of test data are included

  16. Radiological safety of decayed source removal facility (DSRF) - an overview

    International Nuclear Information System (INIS)

    Rajput, Raksha; George, Jain Reji; Pathak, B.K.

    2018-01-01

    Industrial radiography is one of the major applications of radioisotope in engineering industry for Non-Destructive Testing (NDT). The equipment used for this purpose is called Industrial Radiography Exposure Device (IGRED) or radiography (RG) camera. In India, more than 1800 IGREDs including imported cameras are being used in NDT industry. These cameras are of different types and have various capacities to house different radioisotopes. Generally, 192 Ir sources are being used for NDT work. The sources are being supplied by BRIT to the users. After the useful period of the utilization of gamma intensity, the decayed source is returned to BRIT in RG camera. The decayed source is removed from the camera in the Decayed Source Removal Facility (DSRF). This facility serves the purpose of a miniature hot-cell with the capability of storing the decayed sources which are removed from the cameras. The empty camera is inspected for its mechanical functions and sent to BRIT's hot-cell for loading the new source. DSRF is situated at BRIT Vashi Complex. This paper deals with the radiological safety in the operation of DSRF for removing decayed sources from industrial radiography cameras

  17. Test facilities for future linear colliders

    International Nuclear Information System (INIS)

    Ruth, R.D.

    1995-12-01

    During the past several years there has been a tremendous amount of progress on Linear Collider technology world wide. This research has led to the construction of the test facilities described in this report. Some of the facilities will be complete as early as the end of 1996, while others will be finishing up around the end 1997. Even now there are extensive tests ongoing for the enabling technologies for all of the test facilities. At the same time the Linear Collider designs are quite mature now and the SLC is providing the key experience base that can only come from a working collider. All this taken together indicates that the technology and accelerator physics will be ready for a future Linear Collider project to begin in the last half of the 1990s

  18. 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

  19. 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

  20. Kauai Test Facility hazards assessment document

    Energy Technology Data Exchange (ETDEWEB)

    Swihart, A

    1995-05-01

    The Department of Energy Order 55003A requires facility-specific hazards assessment be prepared, maintained, and used for emergency planning purposes. This hazards assessment document describes the chemical and radiological hazards associated with the Kauai Test Facility, Barking Sands, Kauai, Hawaii. The Kauai Test Facility`s chemical and radiological inventories were screened according to potential airborne impact to onsite and offsite individuals. The air dispersion model, ALOHA, estimated pollutant concentrations downwind from the source of a release, taking into consideration the toxicological and physical characteristics of the release site, the atmospheric conditions, and the circumstances of the release. The greatest distance to the Early Severe Health Effects threshold is 4.2 kilometers. The highest emergency classification is a General Emergency at the {open_quotes}Main Complex{close_quotes} and a Site Area Emergency at the Kokole Point Launch Site. The Emergency Planning Zone for the {open_quotes}Main Complex{close_quotes} is 5 kilometers. The Emergency Planning Zone for the Kokole Point Launch Site is the Pacific Missile Range Facility`s site boundary.

  1. Safety analysis of the Los Alamos critical experiments facility

    International Nuclear Information System (INIS)

    Paxton, H.C.

    1975-10-01

    The safety of Pajarito Site critical assembly operations depends upon protection built into the facility, upon knowledgeable personnel, and upon good practice as defined by operating procedures and experimental plans. Distance, supplemented by shielding in some cases, would protect personnel against an extreme accident generating 10 19 fissions. During the facility's 28-year history, the direct cost of criticality accidents has translated to a risk of less than $200 per year

  2. A bicycle safety index for evaluating urban street facilities.

    Science.gov (United States)

    Asadi-Shekari, Zohreh; Moeinaddini, Mehdi; Zaly Shah, Muhammad

    2015-01-01

    The objectives of this research are to conceptualize the Bicycle Safety Index (BSI) that considers all parts of the street and to propose a universal guideline with microscale details. A point system method comparing existing safety facilities to a defined standard is proposed to estimate the BSI. Two streets in Singapore and Malaysia are chosen to examine this model. The majority of previous measurements to evaluate street conditions for cyclists usually cannot cover all parts of streets, including segments and intersections. Previous models also did not consider all safety indicators and cycling facilities at a microlevel in particular. This study introduces a new concept of a practical BSI to complete previous studies using its practical, easy-to-follow, point system-based outputs. This practical model can be used in different urban settings to estimate the level of safety for cycling and suggest some improvements based on the standards.

  3. Comprehensive safety cases for radioactive waste management facilities

    International Nuclear Information System (INIS)

    Woollam, P.B.

    1993-01-01

    Probabilistic safety assessment methodology is being applied by Nuclear Electric plc (NE) to the development of comprehensive safety cases for the radioactive waste management processing and accumulation facilities associated with its 26 reactor systems. This paper describes the methodology and the safety case assessment criteria employed by NE. An overview of the results from facilities used by the first 16 reactors is presented, together with more detail of a specific safety analysis: storage of fuel element debris. No risk to the public greater than 10 -6 /y has been identified and the more significant risks arise from the potential for radioactive waste fires. There are no unacceptable risks from external hazards such as flooding, aircrash or seismic events. Some operations previously expected to have significant risks in fact have negligible risks, while the few faults with risks exceeding the assessment criteria were only identified as a result of this study

  4. Decommissioning of Medical, Industrial and Research Facilities. Safety Guide

    International Nuclear Information System (INIS)

    2010-01-01

    Radioactive waste is produced in the generation of nuclear power and the use of radioactive materials in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized, and considerable experience has been gained in this field. The IAEA's Radioactive Waste Safety Standards Programme aimed at establishing a coherent and comprehensive set of principles and requirements for the safe management of waste and formulating the guidelines necessary for their application. This is accomplished within the IAEA Safety Standards Series in an internally consistent set of publications that reflect an international consensus. The publications will provide Member States with a comprehensive series of internationally agreed publications to assist in the derivation of, and to complement, national criteria, standards and practices. The Safety Standards Series consists of three categories of publications: Safety Fundamentals, Safety Requirements and Safety Guides. With respect to the Radioactive Waste Safety Standards Programme, the set of publications is currently undergoing review to ensure a harmonized approach throughout the Safety Standards Series. This Safety Guide addresses the subject of decommissioning of medical, industrial and research facilities where radioactive materials and sources are produced, received, used and stored. It is intended to provide guidance to national authorities and operating organizations, particularly to those in developing countries (as such facilities are predominant in these countries), for the planning and safe management of the decommissioning of such facilities. The Safety Guide has been prepared through a series of Consultants meetings and a Technical Committee meeting

  5. Decommissioning of medical, industrial and research facilities. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    Radioactive waste is produced in the generation of nuclear power and the use of radioactive materials in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized, and considerable experience has been gained in this field. The IAEA's Radioactive Waste Safety Standards Programme aimed at establishing a coherent and comprehensive set of principles and requirements for the safe management of waste and formulating the guidelines necessary for their application. This is accomplished within the IAEA Safety Standards Series in an internally consistent set of publications that reflect an international consensus. The publications will provide Member States with a comprehensive series of internationally agreed publications to assist in the derivation of, and to complement, national criteria, standards and practices. The Safety Standards Series consists of three categories of publications: Safety Fundamentals, Safety Requirements and Safety Guides. With respect to the Radioactive Waste Safety Standards Programme, the set of publications is currently undergoing review to ensure a harmonized approach throughout the Safety Standards Series. This Safety Guide addresses the subject of decommissioning of medical, industrial and research facilities where radioactive materials and sources are produced, received, used and stored. It is intended to provide guidance to national authorities and operating organizations, particularly to those in developing countries (as such facilities are predominant in these countries), for the planning and safe management of the decommissioning of such facilities. The Safety Guide has been prepared through a series of Consultants meetings and a Technical Committee meeting

  6. Plasma-Materials Interactions Test Facility

    International Nuclear Information System (INIS)

    Uckan, T.

    1986-11-01

    The Plasma-Materials Interactions Test Facility (PMITF), recently designed and constructed at Oak Ridge National Laboratory (ORNL), is an electron cyclotron resonance microwave plasma system with densities around 10 11 cm -3 and electron temperatures of 10-20 eV. The device consists of a mirror cell with high-field-side microwave injection and a heating power of up to 0.8 kW(cw) at 2.45 GHz. The facility will be used for studies of plasma-materials interactions and of particle physics in pump limiters and for development and testing of plasma edge diagnostics

  7. 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)

  8. 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....

  9. 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.

  10. 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)

  11. 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)

  12. 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

  13. 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)

  14. 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)

  15. DeBeNe Test Facilities for Fast Breeder Development

    International Nuclear Information System (INIS)

    Storz, R.

    1980-10-01

    This report gives an overview and a short description of the test facilities constructed and operated within the collaboration for fast breeder development in Germany, Belgium and the Netherlands. The facilities are grouped into Sodium Loops (Large Facilities and Laboratory Loops), Special Equipment including Hot Cells and Reprocessing, Test Facilities without Sodium, Zero Power Facilities and In-pile Loops including Irradiation Facilities

  16. Safety assessment of a borehole type disposal facility using the ISAM methodology

    International Nuclear Information System (INIS)

    Blerk, J.J. van; Yucel, V.; Kozak, M.W.; Moore, B.A.

    2002-01-01

    As part of the IAEA's Co-ordinated Research Project (CRP) on Improving Long-term of Safety Assessment Methodologies for Near Surface Waste Disposal Facilities (ISAM), three example cases were developed. The aim was to test the ISAM safety assessment methodology using as realistic as possible data. One of the Test Cases, the Borehole Test Case (BTC), related to a proposed future disposal option for disused sealed radioactive sources. This paper uses the various steps of the ISAM safety assessment methodology to describe the work undertaken by ISAM participants in developing the BTC and provides some general conclusions that can be drawn from the findings of their work. (author)

  17. Safety aspects of front-end fuel cycle facilities

    International Nuclear Information System (INIS)

    Srinivasan, G.R.

    2003-01-01

    Safety of fuel cycle facilities (FCFs) other than Nuclear Power Plants is gaining importance all over the nuclear world as one would not like to leave behind any area of nuclear field in the journey toward excellence in the safe conduct of business in the whole of the nuclear industry. Safety should be part of every day activities, procedures, business practices, system and in fact of the people themselves

  18. Safety design of the international fusion materials irradiation facility (IFMIF)

    International Nuclear Information System (INIS)

    Konishi, Satoshi; Yamaki, Daiju; Katsuta, Hiroji; Moeslang, Anton; Jameson, R.A.; Martone, Marcello; Shannon, T.E.

    1997-11-01

    In the Conceptual Design Activity of the IFMIF, major subsystems, as well as the entire facility is carefully designed to satisfy the safety requirements for any possible construction sites. Each subsystem is qualitatively analyzed to identify possible hazards to the workers, public and environments using Failure Mode and Effect Analysis (FMEA). The results are reflected in the design and operation procedure. Shielding of radiation, particularly neutron around the test cell is one of the most important issue in normal operation. Radiation due to beam halo and activation is a hazard for operation personnel in the accelerator system. For the maintenance, remote handling technology is designed to be applied in various facilities of the IFMIF. Lithium loop and target system hold the majority of the radioactive material in the facility. Tritium and beryllium-7 are generated by the nuclear reaction during operation and thus needed to be removed continuously. They are also the potential hazards of airborne source in off-normal events. Minimization of inventory, separation and immobilization, and multiple confinement are considered in the design. Generation of radioactive waste is anticipated to be minor, but waste treatment systems for gas, liquid and solid wastes are designed to minimize the environmental impact. Lithium leak followed by a fire is a major concern, and extensive prevention plan is made in the target design. One of the design option considered is composed of; primary enclosure of the lithium loop, secondary containment filled with positive pressure argon, and an air tight lithium cell made of concrete with a steel lining. This study will report some technical issues considered in the design of IFMIF. It was concluded that the IFMIF can be designed and constructed to meet or exceed current safely standards for workers, public and the environment with existing technology and reasonable construction cost. (J.P.N.)

  19. Startup of Large Coil Test Facility

    International Nuclear Information System (INIS)

    Haubenreich, P.N.; Bohanan, R.E.; Fietz, W.A.; Luton, J.N.; May, J.R.

    1984-01-01

    The Large Coil Test Facility (LCTF) is being used to test superconducting toroidal field coils about one-third the size of those for INTOR. Eventually, six different coils from four countries will be tested. Operations began in 1983 with acceptance testing of the helium refrigerator/liquefier system. Comprehensive shakedown of the facility and tests with the first three coils (from Japan, the United States, and Switzerland) were successfully accomplished in the summer of 1984. Currents up to 10,200 A and fields up to 6.4 T were reached. Data were obtained on performance of refrigerator, helium distribution, power supplies, controls, and data acquisition systems and on the acoustic emission, voltages, currents, and mechanical strains during charging and discharging the coils

  20. Startup of Large Coil Test Facility

    International Nuclear Information System (INIS)

    Haubenreich, P.N.; Bohanan, R.E.; Fietz, W.A.; Luton, J.N.; May, J.R.

    1985-01-01

    The Large Coil Test Facility (LCTF) is being used to test superconducting toroidal field coils about one-third the size of those for INTOR. Eventually, six different coils from four countries will be tested. Operations began in 1983 with acceptance testing of the helium refrigerator/liquefier system. Comprehensive shakedown of the facility and tests with the first three coils (from Japan, the United States, and Switzerland) were successfully accomplished in the summer of 1984. Currents up to 10,200 A and fields up to 6.4 T were reached. Data were obtained on performance of refrigerator, helium distribution, power supplies, controls, and data acquisition systems and on the acoustic emission, voltages, currents, and mechanical strains during charging and discharging the coils

  1. Embracing Safe Ground Test Facility Operations and Maintenance

    Science.gov (United States)

    Dunn, Steven C.; Green, Donald R.

    2010-01-01

    Conducting integrated operations and maintenance in wind tunnel ground test facilities requires a balance of meeting due dates, efficient operation, responsiveness to the test customer, data quality, effective maintenance (relating to readiness and reliability), and personnel and facility safety. Safety is non-negotiable, so the balance must be an "and" with other requirements and needs. Pressure to deliver services faster at increasing levels of quality in under-maintained facilities is typical. A challenge for management is to balance the "need for speed" with safety and quality. It s especially important to communicate this balance across the organization - workers, with a desire to perform, can be tempted to cut corners on defined processes to increase speed. Having a lean staff can extend the time required for pre-test preparations, so providing a safe work environment for facility personnel and providing good stewardship for expensive National capabilities can be put at risk by one well-intending person using at-risk behavior. This paper documents a specific, though typical, operational environment and cites management and worker safety initiatives and tools used to provide a safe work environment. Results are presented and clearly show that the work environment is a relatively safe one, though still not good enough to keep from preventing injury. So, the journey to a zero injury work environment - both in measured reality and in the minds of each employee - continues. The intent of this paper is to provide a benchmark for others with operational environments and stimulate additional sharing and discussion on having and keeping a safe work environment.

  2. Use of risk information to safety regulation. Reprocessing facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    A procedure of probabilistic risk assessment (PRA) for a reprocessing facility has been under the development aiming to utilize risk information for safety regulations in this project. Activities in the fiscal year 2012 are summarized in the paper. A major activity is a fundamental study on a concept of serious accidents, requirements of serious accident management, and a policy of utilizing risk information for fabrication and reprocessing facilities. Other than the activity a study on release and transport of aerial radioactive materials at a serious accident in a reprocessing facility has been conducted. The outline and results are provided in the chapter 1 and 2 respectively. (author)

  3. Safety at the End of a Nuclear Facility's Life

    International Nuclear Information System (INIS)

    Geis, John A.; McEahern, Patrice; Evans, Brad

    2004-01-01

    The objective of this paper is to capture the changes that are caused by the transition from nuclear operation through closure of defense nuclear facilities and convey lessons learned from their deactivation, decontamination and demolition. The specific area of discussion is focused on the planned reduction of safety equipment and consequent shift in hazard controls and safety management programs as the facility moves toward closure. The premise of the paper is that as the dominant hazards transition from nuclear to radiological and/or industrial, the facility control of the hazards and response to the potential upset conditions must transition as well to ensure safe and efficient operations. Using recent experience of the accelerated closure mission for U. S. Department of Energy (DOE) defense nuclear facilities at Rocky Flats Environmental Technology Site, the current culture with respect to developing and implementing hazard controls and response to upset conditions is illustrated. Several events have been documented that provide insight into the challenges facing line managers and safety professionals at the end of a facility's life cycle. Replacing permanent systems with temporary equipment challenges the traditional concept of reliability. Workers disassemble safety systems daily, but must rely on some of these components or redundant systems as work continues. Decisions governing upkeep of systems that await demolition balance the risk of running to failure against the cost benefit of maintenance and repair. This is further complicated as regulators and safety professionals are often unfamiliar with these new conditions and continue to view facility work activities and potential upset conditions from a nuclear operations perspective. The results of this paper evaluate the differences in how regulatory, safety basis, and operational practices must adapt to the dynamic environment of decontamination and decommissioning in contrast to the relatively constant

  4. Extreme meteorological events and nuclear facilities safety

    International Nuclear Information System (INIS)

    Almeida, Patricia Moco Princisval

    2006-01-01

    An External Event is an event that originates outside the site and whose effects on the Nuclear Power Plants (NPP) should be considered. Such events could be of natural or human induced origin and should be identified and selected for design purposes during the site evaluation process. This work shows that the subtropics and mid latitudes of South America east of the Andes Mountain Range have been recognized as prone to severe convective weather. In Brazil, the events of tornadoes are becoming frequent; however there is no institutionalized procedure for a systematic documentation of severe weather. The information is done only for some scientists and by the newspapers. Like strong wind can affect the structural integrity of buildings or the pressure differential can affect the ventilation system, our concern is the safety of NPP and for this purpose the recommendations of International Atomic Energy Agency, Nuclear Regulatory Commission and Comissao Nacional de Energia Nuclear are showed and also a data base of tornadoes in Brazil is done. (author)

  5. Safety Assessment for Facilities and Activities. General Safety Requirements. Pt. 4

    International Nuclear Information System (INIS)

    2009-01-01

    The Safety Fundamentals publication, Fundamental Safety Principles, establishes principles for ensuring the protection of workers, the public and the environment, now and in the future, from harmful effects of ionizing radiation. The objective of this Safety Requirements publication is to establish the generally applicable requirements to be fulfilled in safety assessment for facilities and activities, with special attention paid to defence in depth, quantitative analyses and the application of a graded approach to the ranges of facilities and of activities that are addressed. The publication also addresses the independent verification of the safety assessment that needs to be carried out by the originators and users of the safety assessment. This publication is intended to provide a consistent and coherent basis for safety assessment across all facilities and activities, which will facilitate the transfer of good practices between organizations conducting safety assessments and will assist in enhancing the confidence of all interested parties that an adequate level of safety has been achieved for facilities and activities. The requirements, which are derived from the Fundamental Safety Principles, relate to any human activity that may cause people to be exposed to radiation risks arising from facilities and activities, as follows: Facilities includes: (a) Nuclear power plants; (b) Other reactors (such as research reactors and critical assemblies); (c) Enrichment facilities and fuel fabrication facilities; (d) Conversion facilities used to generate UF6; (e) Storage and reprocessing plants for irradiated fuel; (f) Facilities for radioactive waste management where radioactive waste is treated, conditioned, stored or disposed of; (g) Any other places where radioactive materials are produced, processed, used, handled or stored; (h) Irradiation facilities for medical, industrial, research and other purposes, and any places where radiation generators are installed; (i

  6. Safety assessment for facilities and activities. General safety requirements. Pt. 4

    International Nuclear Information System (INIS)

    2009-01-01

    The Safety Fundamentals publication, Fundamental Safety Principles, establishes principles for ensuring the protection of workers, the public and the environment, now and in the future, from harmful effects of ionizing radiation. The objective of this Safety Requirements publication is to establish the generally applicable requirements to be fulfilled in safety assessment for facilities and activities, with special attention paid to defence in depth, quantitative analyses and the application of a graded approach to the ranges of facilities and of activities that are addressed. The publication also addresses the independent verification of the safety assessment that needs to be carried out by the originators and users of the safety assessment. This publication is intended to provide a consistent and coherent basis for safety assessment across all facilities and activities, which will facilitate the transfer of good practices between organizations conducting safety assessments and will assist in enhancing the confidence of all interested parties that an adequate level of safety has been achieved for facilities and activities. The requirements, which are derived from the Fundamental Safety Principles, relate to any human activity that may cause people to be exposed to radiation risks arising from facilities and activities, as follows: Facilities includes: (a) Nuclear power plants; (b) Other reactors (such as research reactors and critical assemblies); (c) Enrichment facilities and fuel fabrication facilities; (d) Conversion facilities used to generate UF 6 ; (e) Storage and reprocessing plants for irradiated fuel; (f) Facilities for radioactive waste management where radioactive waste is treated, conditioned, stored or disposed of; (g) Any other places where radioactive materials are produced, processed, used, handled or stored; (h) Irradiation facilities for medical, industrial, research and other purposes, and any places where radiation generators are installed; (i

  7. Safety Assessment for Facilities and Activities. General Safety Requirements. Pt. 4

    International Nuclear Information System (INIS)

    2010-01-01

    The Safety Fundamentals publication, Fundamental Safety Principles, establishes principles for ensuring the protection of workers, the public and the environment, now and in the future, from harmful effects of ionizing radiation. The objective of this Safety Requirements publication is to establish the generally applicable requirements to be fulfilled in safety assessment for facilities and activities, with special attention paid to defence in depth, quantitative analyses and the application of a graded approach to the ranges of facilities and of activities that are addressed. The publication also addresses the independent verification of the safety assessment that needs to be carried out by the originators and users of the safety assessment. This publication is intended to provide a consistent and coherent basis for safety assessment across all facilities and activities, which will facilitate the transfer of good practices between organizations conducting safety assessments and will assist in enhancing the confidence of all interested parties that an adequate level of safety has been achieved for facilities and activities. The requirements, which are derived from the Fundamental Safety Principles, relate to any human activity that may cause people to be exposed to radiation risks arising from facilities and activities, as follows: Facilities includes: (a) Nuclear power plants; (b) Other reactors (such as research reactors and critical assemblies); (c) Enrichment facilities and fuel fabrication facilities; (d) Conversion facilities used to generate UF6; (e) Storage and reprocessing plants for irradiated fuel; (f) Facilities for radioactive waste management where radioactive waste is treated, conditioned, stored or disposed of; (g) Any other places where radioactive materials are produced, processed, used, handled or stored; (h) Irradiation facilities for medical, industrial, research and other purposes, and any places where radiation generators are installed; (i

  8. Safety Assessment for Facilities and Activities. General Safety Requirements. Pt. 4

    International Nuclear Information System (INIS)

    2009-01-01

    The Safety Fundamentals publication, Fundamental Safety Principles, establishes principles for ensuring the protection of workers, the public and the environment, now and in the future, from harmful effects of ionizing radiation.? read more The objective of this Safety Requirements publication is to establish the generally applicable requirements to be fulfilled in safety assessment for facilities and activities, with special attention paid to defence in depth, quantitative analyses and the application of a graded approach to the ranges of facilities and of activities that are addressed. The publication also addresses the independent verification of the safety assessment that needs to be carried out by the originators and users of the safety assessment. This publication is intended to provide a consistent and coherent basis for safety assessment across all facilities and activities, which will facilitate the transfer of good practices between organizations conducting safety assessments and will assist in enhancing the confidence of all interested parties that an adequate level of safety has been achieved for facilities and activities. The requirements, which are derived from the Fundamental Safety Principles, relate to any human activity that may cause people to be exposed to radiation risks arising from facilities and activities, as follows: Facilities includes: (a) Nuclear power plants; (b) Other reactors (such as research reactors and critical assemblies); (c) Enrichment facilities and fuel fabrication facilities; (d) Conversion facilities used to generate UF6; (e) Storage and reprocessing plants for irradiated fuel; (f) Facilities for radioactive waste management where radioactive waste is treated, conditioned, stored or disposed of; (g) Any other places where radioactive materials are produced, processed, used, handled or stored; (h) Irradiation facilities for medical, industrial, research and other purposes, and any places where radiation generators are

  9. The advanced neutron source facility: Safety philosophy and studies

    International Nuclear Information System (INIS)

    Greene, S.R.; Harrington, R.M.

    1988-01-01

    The Advanced Neutron Source (ANS) is currently the only new civilian nuclear reactor facility proposed for construction in the United States. Even though the thermal power of this research-oriented reactor is a relatively low 300 MW, the design will undoubtedly receive intense scrutiny before construction is allowed to proceed. Safety studies are already under way to ensure that the maximum degree of safety in incorporated into the design and that the design is acceptable to the Department of Energy (DOE) and can meet the Nuclear Regulatory Commission regulations. This document discusses these safety studies

  10. The BNL Accelerator Test Facility control system

    International Nuclear Information System (INIS)

    Malone, R.; Bottke, I.; Fernow, R.; Ben-Zvi, I.

    1993-01-01

    Described is the VAX/CAMAC-based control system for Brookhaven National Laboratory's Accelerator Test Facility, a laser/linac research complex. Details of hardware and software configurations are presented along with experiences of using Vsystem, a commercial control system package

  11. FFTF [Fast Flux Test Facility] management

    International Nuclear Information System (INIS)

    Bennett, C.L.

    1986-11-01

    Fuel Management at the Fast Flux Test Facility (FFTF) involves more than just the usual ex-core and in-core management of standard fuel and non-fuel components between storage locations and within the core since it is primarily an irradiation test facility. This mission involves testing an ever increasing variety of fueled and non-fueled experiments, each having unique requirements on the reactor core as well as having its own individual impact on the reload design. This paper describes the fuel management process used by the Westinghouse Hanford Company Core Engineering group that has led to the successful reload design of nine operating cycles and the irradiation of over 120 tests

  12. The NRU blowdown test facility commissioning program

    Energy Technology Data Exchange (ETDEWEB)

    Walsworth, J A; Zanatta, R J; Yamazaki, A R; Semeniuk, D D; Wong, W; Dickson, L W; Ferris, C E; Burton, D H [Atomic Energy of Canada Ltd., Chalk River, ON (Canada). Chalk River Nuclear Labs.

    1990-12-31

    A major experimental program has been established at the Chalk River Nuclear Laboratories (CRL) that will provide essential data on the thermal and mechanical behaviour of nuclear fuel under abnormal reactor operating conditions and on the transient release, transport and deposition of fission product activity from severely degraded fuel. A number of severe fuel damage (SFD) experiments will be conducted within the Blowdown Test Facility (BTF) at CRL. A series of experiments are being conducted to commission this new facility prior to the SFD program. This paper describes the features and the commissioning program for the BTF. A development and testing program is described for critical components used on the reactor test section. In-reactor commissioning with a fuel assembly simulator commenced in 1989 June and preliminary results are given. The paper also outlines plans for future all-effects, in-reactor tests of CANDU-designed fuel. (author). 11 refs., 3 tabs., 7 figs.

  13. Conceptual studies of plasma engineering test facility

    International Nuclear Information System (INIS)

    Hiraoka, Toru; Tazima, Teruhiko; Sugihara, Masayoshi; Kasai, Masao; Shinya, Kichiro

    1979-04-01

    Conceptual studies have been made of a Plasma Engineering Test Facility, which is to be constructed following JT-60 prior to the experimental power reactor. The physical aim of this machine is to examine self-ignition conditions. This machine possesses all essential technologies for reactor plasma, i.e. superconducting magnet, remote maintenance, shielding, blanket test modules, tritium handling. Emphasis in the conceptual studies was on structural consistency of the machine and whether the machine would be constructed practically. (author)

  14. ACIGA's high optical power test facility

    International Nuclear Information System (INIS)

    Ju, L; Aoun, M; Barriga, P

    2004-01-01

    Advanced laser interferometer detectors utilizing more than 100 W of laser power and with ∼10 6 W circulating laser power present many technological problems. The Australian Consortium for Interferometric Gravitational Astronomy (ACIGA) is developing a high power research facility in Gingin, north of Perth, Western Australia, which will test techniques for the next generation interferometers. In particular it will test thermal lensing compensation and control strategies for optical cavities in which optical spring effects and parametric instabilities may present major difficulties

  15. Test facility TIMO for testing the ITER model cryopump

    International Nuclear Information System (INIS)

    Haas, H.; Day, C.; Mack, A.; Methe, S.; Boissin, J.C.; Schummer, P.; Murdoch, D.K.

    2001-01-01

    Within the framework of the European Fusion Technology Programme, FZK is involved in the research and development process for a vacuum pump system of a future fusion reactor. As a result of these activities, the concept and the necessary requirements for the primary vacuum system of the ITER fusion reactor were defined. Continuing that development process, FZK has been preparing the test facility TIMO (Test facility for ITER Model pump) since 1996. This test facility provides for testing a cryopump all needed infrastructure as for example a process gas supply including a metering system, a test vessel, the cryogenic supply for the different temperature levels and a gas analysing system. For manufacturing the ITER model pump an order was given to the company L' Air Liquide in the form of a NET contract. (author)

  16. Test facility TIMO for testing the ITER model cryopump

    International Nuclear Information System (INIS)

    Haas, H.; Day, C.; Mack, A.; Methe, S.; Boissin, J.C.; Schummer, P.; Murdoch, D.K.

    1999-01-01

    Within the framework of the European Fusion Technology Programme, FZK is involved in the research and development process for a vacuum pump system of a future fusion reactor. As a result of these activities, the concept and the necessary requirements for the primary vacuum system of the ITER fusion reactor were defined. Continuing that development process, FZK has been preparing the test facility TIMO (Test facility for ITER Model pump) since 1996. This test facility provides for testing a cryopump all needed infrastructure as for example a process gas supply including a metering system, a test vessel, the cryogenic supply for the different temperature levels and a gas analysing system. For manufacturing the ITER model pump an order was given to the company L'Air Liquide in the form of a NET contract. (author)

  17. Technical Safety Requirements for the Waste Storage Facilities

    International Nuclear Information System (INIS)

    Larson, H L

    2007-01-01

    This document contains Technical Safety Requirements (TSR) for the Radioactive and Hazardous Waste Management (RHWM) WASTE STORAGE FACILITIES, which include Area 612 (A612) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area at Lawrence Livermore National Laboratory (LLNL). The TSRs constitute requirements regarding the safe operation of the WASTE STORAGE FACILITIES. These TSRs are derived from the Documented Safety Analysis for the Waste Storage Facilities (DSA) (LLNL 2006). The analysis presented therein determined that the WASTE STORAGE FACILITIES are low-chemical hazard, Hazard Category 2 non-reactor nuclear facilities. The TSRs consist primarily of inventory limits and controls to preserve the underlying assumptions in the hazard and accident analyses. Further, appropriate commitments to safety programs are presented in the administrative controls sections of the TSRs. The WASTE STORAGE FACILITIES are used by RHWM to handle and store hazardous waste, TRANSURANIC (TRU) WASTE, LOW-LEVEL WASTE (LLW), mixed waste, California combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL as well as small amounts from other U.S. Department of Energy (DOE) facilities, as described in the DSA. In addition, several minor treatments (e.g., drum crushing, size reduction, and decontamination) are carried out in these facilities. The WASTE STORAGE FACILITIES are located in two portions of the LLNL main site. A612 is located in the southeast quadrant of LLNL. The A612 fenceline is approximately 220 m west of Greenville Road. The DWTF Storage Area, which includes Building 693 (B693), Building 696 Radioactive Waste Storage Area (B696R), and associated yard areas and storage areas within the yard, is located in the northeast quadrant of LLNL in the DWTF complex. The DWTF Storage Area fenceline is approximately 90 m west of Greenville Road. A612 and the DWTF Storage Area are subdivided into various facilities and storage

  18. Technical Safety Requirements for the Waste Storage Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Larson, H L

    2007-09-07

    This document contains Technical Safety Requirements (TSR) for the Radioactive and Hazardous Waste Management (RHWM) WASTE STORAGE FACILITIES, which include Area 612 (A612) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area at Lawrence Livermore National Laboratory (LLNL). The TSRs constitute requirements regarding the safe operation of the WASTE STORAGE FACILITIES. These TSRs are derived from the Documented Safety Analysis for the Waste Storage Facilities (DSA) (LLNL 2006). The analysis presented therein determined that the WASTE STORAGE FACILITIES are low-chemical hazard, Hazard Category 2 non-reactor nuclear facilities. The TSRs consist primarily of inventory limits and controls to preserve the underlying assumptions in the hazard and accident analyses. Further, appropriate commitments to safety programs are presented in the administrative controls sections of the TSRs. The WASTE STORAGE FACILITIES are used by RHWM to handle and store hazardous waste, TRANSURANIC (TRU) WASTE, LOW-LEVEL WASTE (LLW), mixed waste, California combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL as well as small amounts from other U.S. Department of Energy (DOE) facilities, as described in the DSA. In addition, several minor treatments (e.g., drum crushing, size reduction, and decontamination) are carried out in these facilities. The WASTE STORAGE FACILITIES are located in two portions of the LLNL main site. A612 is located in the southeast quadrant of LLNL. The A612 fenceline is approximately 220 m west of Greenville Road. The DWTF Storage Area, which includes Building 693 (B693), Building 696 Radioactive Waste Storage Area (B696R), and associated yard areas and storage areas within the yard, is located in the northeast quadrant of LLNL in the DWTF complex. The DWTF Storage Area fenceline is approximately 90 m west of Greenville Road. A612 and the DWTF Storage Area are subdivided into various facilities and storage

  19. Fuel supply shutdown facility interim operational safety requirements

    International Nuclear Information System (INIS)

    Besser, R.L.; Brehm, J.R.; Benecke, M.W.; Remaize, J.A.

    1995-01-01

    These Interim Operational Safety Requirements (IOSR) for the Fuel Supply Shutdown (FSS) facility define acceptable conditions, safe boundaries, bases thereof, and management or administrative controls to ensure safe operation. The IOSRs apply to the fuel material storage buildings in various modes (operation, storage, surveillance)

  20. Criticality safety training at the Hot Fuel Examination Facility

    International Nuclear Information System (INIS)

    Garcia, A.S.; Courtney, J.C.; Thelen, V.N.

    1983-01-01

    HFEF comprises four hot cells and out-of-cell support facilities for the US breeder program. The HFEF criticality safety program includes training in the basic theory of criticality and in specific criticality hazard control rules that apply to HFEF. A professional staff-member oversees the implementation of the criticality prevention program

  1. Transuranic waste storage and assay facility (TRUSAF) interim safety basis

    International Nuclear Information System (INIS)

    Gibson, K.D.

    1995-09-01

    The TRUSAF ISB is based upon current facility configuration and procedures. The purpose of the document is to provide the basis for interim operation or restrictions on interim operations and the authorization basis for the TRUSAF at the Hanford Site. The previous safety analysis document TRUSAF hazards Identification and Evaluation (WHC 1977) is superseded by this document

  2. The study on safety facility criteria for radioactive waste repository

    International Nuclear Information System (INIS)

    Lee, S. H.; Choi, M. H.; Han, S. H. and others

    1992-12-01

    The radioactive waste repository are necessary to install the engineered safety systems to secure the safety for operation of the repository in the event of fire and earthquake. Since the development of safety facility criteria requires a thorough understanding about the characteristics of the engineered safety systems, we should investigate by means of literature survey and visit SKB. In particular, definition, composition of the systems, functional requirement of the systems, engineered safety systems of foreign countries, system design, operation and maintenance requirement should be investigated : fire protection system, ventilation system, drainage system, I and C system, electric system, radiation monitoring system. This proposed criteria consist of purpose, scope of application, ventilation system, fire protection system, drainage system, electric system and this proposed criteria can be applied as a basic reference for the final criteria

  3. Hot helium flow test facility summary report

    International Nuclear Information System (INIS)

    1980-06-01

    This report summarizes the results of a study conducted to assess the feasibility and cost of modifying an existing circulator test facility (CTF) at General Atomic Company (GA). The CTF originally was built to test the Delmarva Power and Light Co. steam-driven circulator. This circulator, as modified, could provide a source of hot, pressurized helium for high-temperature gas-cooled reactor (HTGR) and gas-cooled fast breeder reactor (GCFR) component testing. To achieve this purpose, a high-temperature impeller would be installed on the existing machine. The projected range of tests which could be conducted for the project is also presented, along with corresponding cost considerations

  4. 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

  5. Safety in Elevators and Grain Handling Facilities. Module SH-27. Safety and Health.

    Science.gov (United States)

    Center for Occupational Research and Development, Inc., Waco, TX.

    This student module on safety in elevators and grain handling facilities is one of 50 modules concerned with job safety and health. Following the introduction, 15 objectives (each keyed to a page in the text) the student is expected to accomplish are listed (e.g., Explain how explosion suppression works). Then each objective is taught in detail,…

  6. The WR-1 corrosion test facility

    International Nuclear Information System (INIS)

    Murphy, E.V.; Simmons, G.R.

    1978-07-01

    This report describes a new Corrosion Test Facility which has recently been installed in the WR-1 organic-cooled research reactor. The irradiation facility is a single insert, installed in a reactor site, which can deliver a fast neutron flux density of 2.65 x 10 17 neutrons/(m 2 .s) to specimens under irradiation. A self-contained controlled-chemistry cooling water circuit removes the gamma- and neutron-heat generated in the insert and specimens. Specimen temperatures typically vary from 245 deg C to 280 deg C across the insert core region. (author)

  7. Psychometric model for safety culture assessment in nuclear research facilities

    International Nuclear Information System (INIS)

    Nascimento, C.S. do; Andrade, D.A.; Mesquita, R.N. de

    2017-01-01

    Highlights: • A psychometric model to evaluate ‘safety climate’ at nuclear research facilities. • The model presented evidences of good psychometric qualities. • The model was applied to nuclear research facilities in Brazil. • Some ‘safety culture’ weaknesses were detected in the assessed organization. • A potential tool to develop safety management programs in nuclear facilities. - Abstract: A safe and reliable operation of nuclear power plants depends not only on technical performance, but also on the people and on the organization. Organizational factors have been recognized as the main causal mechanisms of accidents by research organizations through USA, Europe and Japan. Deficiencies related with these factors reveal weaknesses in the organization’s safety culture. A significant number of instruments to assess the safety culture based on psychometric models that evaluate safety climate through questionnaires, and which are based on reliability and validity evidences, have been published in health and ‘safety at work’ areas. However, there are few safety culture assessment instruments with these characteristics (reliability and validity) available on nuclear literature. Therefore, this work proposes an instrument to evaluate, with valid and reliable measures, the safety climate of nuclear research facilities. The instrument was developed based on methodological principles applied to research modeling and its psychometric properties were evaluated by a reliability analysis and validation of content, face and construct. The instrument was applied to an important nuclear research organization in Brazil. This organization comprises 4 research reactors and many nuclear laboratories. The survey results made possible a demographic characterization and the identification of some possible safety culture weaknesses and pointing out potential areas to be improved in the assessed organization. Good evidence of reliability with Cronbach's alpha

  8. Psychometric model for safety culture assessment in nuclear research facilities

    Energy Technology Data Exchange (ETDEWEB)

    Nascimento, C.S. do, E-mail: claudio.souza@ctmsp.mar.mil.br [Centro Tecnológico da Marinha em São Paulo (CTMSP), Av. Professor Lineu Prestes 2468, 05508-000 São Paulo, SP (Brazil); Andrade, D.A., E-mail: delvonei@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN – SP), Av. Professor Lineu Prestes 2242, 05508-000 São Paulo, SP (Brazil); Mesquita, R.N. de, E-mail: rnavarro@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN – SP), Av. Professor Lineu Prestes 2242, 05508-000 São Paulo, SP (Brazil)

    2017-04-01

    Highlights: • A psychometric model to evaluate ‘safety climate’ at nuclear research facilities. • The model presented evidences of good psychometric qualities. • The model was applied to nuclear research facilities in Brazil. • Some ‘safety culture’ weaknesses were detected in the assessed organization. • A potential tool to develop safety management programs in nuclear facilities. - Abstract: A safe and reliable operation of nuclear power plants depends not only on technical performance, but also on the people and on the organization. Organizational factors have been recognized as the main causal mechanisms of accidents by research organizations through USA, Europe and Japan. Deficiencies related with these factors reveal weaknesses in the organization’s safety culture. A significant number of instruments to assess the safety culture based on psychometric models that evaluate safety climate through questionnaires, and which are based on reliability and validity evidences, have been published in health and ‘safety at work’ areas. However, there are few safety culture assessment instruments with these characteristics (reliability and validity) available on nuclear literature. Therefore, this work proposes an instrument to evaluate, with valid and reliable measures, the safety climate of nuclear research facilities. The instrument was developed based on methodological principles applied to research modeling and its psychometric properties were evaluated by a reliability analysis and validation of content, face and construct. The instrument was applied to an important nuclear research organization in Brazil. This organization comprises 4 research reactors and many nuclear laboratories. The survey results made possible a demographic characterization and the identification of some possible safety culture weaknesses and pointing out potential areas to be improved in the assessed organization. Good evidence of reliability with Cronbach's alpha

  9. NIF conventional facilities construction health and safety plan

    International Nuclear Information System (INIS)

    Benjamin, D W

    1998-01-01

    The purpose of this Plan is to outline the minimum health and safety requirements to which all participating Lawrence Livermore National Laboratory (LLNL) and non-LLNL employees (excluding National Ignition Facility [NIF] specific contractors and subcontractors covered under the construction subcontract packages (e.g., CSP-9)-see Construction Safety Program for the National Ignition Facility [CSP] Section I.B. ''NIF Construction Contractors and Subcontractors'' for specifics) shall adhere to for preventing job-related injuries and illnesses during Conventional Facilities construction activities at the NIF Project. For the purpose of this Plan, the term ''LLNL and non-LLNL employees'' includes LLNL employees, LLNL Plant Operations staff and their contractors, supplemental labor, contract labor, labor-only contractors, vendors, DOE representatives, personnel matrixed/assigned from other National Laboratories, participating guests, and others such as visitors, students, consultants etc., performing on-site work or services in support of the NIF Project. Based upon an activity level determination explained in Section 1.2.18, in this document, these organizations or individuals may be required by site management to prepare their own NIF site-specific safety plan. LLNL employees will normally not be expected to prepare a site-specific safety plan. This Plan also outlines job-specific exposures and construction site safety activities with which LLNL and non-LLNL employees shall comply

  10. Management concepts and safety applications for nuclear fuel facilities

    International Nuclear Information System (INIS)

    Eisner, H.; Scotti, R.S.

    1995-05-01

    This report presents an overview of effectiveness of management control of safety. It reviews several modern management control theories as well as the general functions of management and relates them to safety issues at the corporate and at the process safety management (PSM) program level. Following these discussions, structured technique for assessing management of the safety function is suggested. Seven modern management control theories are summarized, including business process reengineering, the learning organization, capability maturity, total quality management, quality assurance and control, reliability centered maintenance, and industrial process safety. Each of these theories is examined for-its principal characteristics and implications for safety management. The five general management functions of planning, organizing, directing, monitoring, and integrating, which together provide control over all company operations, are discussed. Under the broad categories of Safety Culture, Leadership and Commitment, and Operating Excellence, key corporate safety elements and their subelements are examined. The three categories under which PSM program-level safety issues are described are Technology, Personnel, and Facilities

  11. Management concepts and safety applications for nuclear fuel facilities

    Energy Technology Data Exchange (ETDEWEB)

    Eisner, H.; Scotti, R.S. [George Washington Univ., Washington, DC (United States). School of Engineering and Applied Science; Delicate, W.S. [KEVRIC Co., Inc., Silver Spring, MD (United States)

    1995-05-01

    This report presents an overview of effectiveness of management control of safety. It reviews several modern management control theories as well as the general functions of management and relates them to safety issues at the corporate and at the process safety management (PSM) program level. Following these discussions, structured technique for assessing management of the safety function is suggested. Seven modern management control theories are summarized, including business process reengineering, the learning organization, capability maturity, total quality management, quality assurance and control, reliability centered maintenance, and industrial process safety. Each of these theories is examined for-its principal characteristics and implications for safety management. The five general management functions of planning, organizing, directing, monitoring, and integrating, which together provide control over all company operations, are discussed. Under the broad categories of Safety Culture, Leadership and Commitment, and Operating Excellence, key corporate safety elements and their subelements are examined. The three categories under which PSM program-level safety issues are described are Technology, Personnel, and Facilities.

  12. Status of safety at Areva group facilities. 2007 annual report

    International Nuclear Information System (INIS)

    2007-01-01

    This report describes the status of nuclear safety and radiation protection in the facilities of the AREVA group and gives information on radiation protection in the service operations, as observed through the inspection programs and analyses carried out by the General Inspectorate in 2007. Having been submitted to the group's Supervisory Board, this report is sent to the bodies representing the personnel. Content: 1 - A look back at 2007 by the AREVA General Inspector: Visible progress in 2007, Implementation of the Nuclear Safety Charter, Notable events; 2 - Status of nuclear safety and radiation protection in the nuclear facilities and service operations: Personnel radiation protection, Event tracking, Service operations, Criticality control, Radioactive waste and effluent management; 3 - Performance improvement actions; 4 - Description of the General Inspectorate; 5 - Glossary

  13. Several new thermo-hydraulic test facilities in NPIC

    International Nuclear Information System (INIS)

    Ye Shurong; Sun Yufa; Ji Fuyun; Zong Guifang; Guo Zhongchuan

    1997-01-01

    Several new thermo-hydraulic test facilities are under construction in Nuclear Power Institute of Chinese (NPIC) at Chengdu. These facilities include: 1. Nuclear Power Component Comprehensive Test Facility. 2. Reactor Hydraulic Modeling Test Facility. 3. Control Rod Drive Line Hydraulic Test Facility. 4. Large Scale Thermo-Hydraulic Test Facility. The construction of these facilities will make huge progress in the research and development capability of nuclear power technology in CHINA. The author will present a brief description of the design parameters flowchart and test program of these facilities

  14. 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

  15. Safety study of fire protection for nuclear fuel cycle facility

    International Nuclear Information System (INIS)

    2013-01-01

    Insufficiencies in the fire protection system of the nuclear reactor facilities were pointed out when the fire occurred due to the Niigata prefecture-Chuetsu-oki Earthquake in July, 2007. This prompted the revision of the fire protection safety examination guideline for nuclear reactors as well as commercial guidelines. The commercial guidelines have been endorsed by the regulatory body. Now commercial fire protection standards for nuclear facilities such as the design guideline and the management guideline for protecting fire in the Light Water Reactors (LWRs) are available, however, those to apply to the nuclear fuel cycle facilities such as mixed oxide fuel fabrication facility (MFFF) have not been established. For the improvement of fire protection system of the nuclear fuel cycle facilities, the development of a standard for the fire protection, corresponding to the commercial standard for LWRs were required. Thus, Japan Nuclear Energy Safety Organization (JNES) formulated a fire protection guidelines for nuclear fuel cycle facilities as a standard relevant to the fire protection of the nuclear fuel cycle facilities considering functions specific to the nuclear fuel cycle facilities. In formulating the guidelines, investigation has been conduced on the commercial guidelines for nuclear reactors in Japan and the standards relevant to the fire protection of nuclear facilities in USA and other countries as well as non-nuclear industrial fire protection standards. The guideline consists of two parts; Equipments and Management, as the commercial guidances of the nuclear reactor. In addition, the acquisition of fire evaluation data for a components (an electric cabinet, cable, oil etc.) targeted for spread of fire and the evaluation model of fire source were continued for the fire hazard analysis (FHA). (author)

  16. Mirror Fusion Test Facility magnet system

    International Nuclear Information System (INIS)

    VanSant, J.H.; Kozman, T.A.; Bulmer, R.H.; Ng, D.S.

    1981-01-01

    In 1979, R.H. Bulmer of Lawrence Livermore National Laboratory (LLNL) discussed a proposed tandem-mirror magnet system for the Mirror Fusion Test Facility (MFTF) at the 8th symposium on Engineering Problems in Fusion Research. Since then, Congress has voted funds for expanding LLNL's MFTF to a tandem-mirror facility (designated MFTF-B). The new facility, scheduled for completion by 1985, will seek to achieve two goals: (1) Energy break-even capability (Q or the ratio of fusion energy to plasma heating energy = 1) of mirror fusion, (2) Engineering feasibility of reactor-scale machines. Briefly stated, 22 superconducting magnets contained in a 11-m-diam by 65-m-long vacuum vessel will confine a fusion plasma fueled by 80 axial streaming-plasma guns and over 40 radial neutral beams. We have already completed a preliminary design of this magnet system

  17. Documented Safety Analysis for the Waste Storage Facilities March 2010

    Energy Technology Data Exchange (ETDEWEB)

    Laycak, D T

    2010-03-05

    This Documented Safety Analysis (DSA) for the Waste Storage Facilities was developed in accordance with 10 CFR 830, Subpart B, 'Safety Basis Requirements,' and utilizes the methodology outlined in DOE-STD-3009-94, Change Notice 3. The Waste Storage Facilities consist of Area 625 (A625) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area portion of the DWTF complex. These two areas are combined into a single DSA, as their functions as storage for radioactive and hazardous waste are essentially identical. The B695 Segment of DWTF is addressed under a separate DSA. This DSA provides a description of the Waste Storage Facilities and the operations conducted therein; identification of hazards; analyses of the hazards, including inventories, bounding releases, consequences, and conclusions; and programmatic elements that describe the current capacity for safe operations. The mission of the Waste Storage Facilities is to safely handle, store, and treat hazardous waste, transuranic (TRU) waste, low-level waste (LLW), mixed waste, combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL (as well as small amounts from other DOE facilities).

  18. Documented Safety Analysis for the Waste Storage Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Laycak, D

    2008-06-16

    This documented safety analysis (DSA) for the Waste Storage Facilities was developed in accordance with 10 CFR 830, Subpart B, 'Safety Basis Requirements', and utilizes the methodology outlined in DOE-STD-3009-94, Change Notice 3. The Waste Storage Facilities consist of Area 625 (A625) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area portion of the DWTF complex. These two areas are combined into a single DSA, as their functions as storage for radioactive and hazardous waste are essentially identical. The B695 Segment of DWTF is addressed under a separate DSA. This DSA provides a description of the Waste Storage Facilities and the operations conducted therein; identification of hazards; analyses of the hazards, including inventories, bounding releases, consequences, and conclusions; and programmatic elements that describe the current capacity for safe operations. The mission of the Waste Storage Facilities is to safely handle, store, and treat hazardous waste, transuranic (TRU) waste, low-level waste (LLW), mixed waste, combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL (as well as small amounts from other DOE facilities).

  19. Cold moderator test facilities working group

    International Nuclear Information System (INIS)

    Bauer, Guenter S.; Lucas, A. T.

    1997-09-01

    The working group meeting was chaired by Bauer and Lucas.Testing is a vital part of any cold source development project. This applies to specific physics concept verification, benchmarking in conjunction with computer modeling and engineering testing to confirm the functional viability of a proposed system. Irradiation testing of materials will always be needed to continuously extend a comprehensive and reliable information database. An ever increasing worldwide effort to enhance the performance of reactor and accelerator based neutron sources, coupled with the complexity and rising cost of building new generation facilities, gives a new dimension to cold source development and testing programs. A stronger focus is now being placed on the fine-tuning of cold source design to maximize its effectiveness in fully exploiting the facility. In this context, pulsed spallation neutron sources pose an extra challenge due to requirements regarding pulse width and shape which result from a large variety of different instrument concepts. The working group reviewed these requirements in terms of their consequences on the needs for testing equipment and compiled a list of existing and proposed facilities suitable to carry out the necessary development work.

  20. HTS power lead testing at the Fermilab magnet test facility

    Energy Technology Data Exchange (ETDEWEB)

    Rabehl, R.; Carcagno, R.; Feher, S.; Huang, Y.; Orris, D.; Pischalnikov, Y.; Sylvester, C.; Tartaglia, M.; /Fermilab

    2005-08-01

    The Fermilab Magnet Test Facility has tested high-temperature superconductor (HTS) power leads for cryogenic feed boxes to be placed at the Large Hadron Collider (LHC) interaction regions and at the new BTeV C0 interaction region of the Fermilab Tevatron. A new test facility was designed and operated, successfully testing 20 pairs of HTS power leads for the LHC and 2 pairs of HTS power leads for the BTeV experiment. This paper describes the design and operation of the cryogenics, process controls, data acquisition, and quench management systems. Results from the facility commissioning are included, as is the performance of a new insulation method to prevent frost accumulation on the warm ends of the power leads.

  1. HTS power lead testing at the Fermilab magnet test facility

    International Nuclear Information System (INIS)

    Rabehl, R.; Carcagno, R.; Feher, S.; Huang, Y.; Orris, D.; Pischalnikov, Y.; Sylvester, C.; Tartaglia, M.

    2005-01-01

    The Fermilab Magnet Test Facility has tested high-temperature superconductor (HTS) power leads for cryogenic feed boxes to be placed at the Large Hadron Collider (LHC) interaction regions and at the new BTeV CO interaction region of the Fermilab Tevatron. A new test facility was designed and operated, successfully testing 20 pairs of HTS power leads for the LHC and 2 pairs of HTS power leads for the BTeV experiment. This paper describes the design and operation of the cryogenics, process controls, data acquisition, and quench management systems. Results from the facility commissioning are included, as is the performance of a new insulation method to prevent frost accumulation on the warm ends of the power leads

  2. 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)

  3. Radiation Safety of Gamma, Electron and X Ray Irradiation Facilities. Specific Safety Guide (Spanish Edition)

    International Nuclear Information System (INIS)

    2015-01-01

    The objective of this Safety Guide is to provide recommendations on how to meet the requirements of the BSS with regard to irradiation facilities. This Safety Guide provides specific, practical recommendations on the safe design and operation of gamma, electron and X ray irradiators for use by operating organizations and the designers of these facilities, and by regulatory bodies. SCOPE. The facilities considered in this publication include five types of irradiator, whether operated on a commercial basis or for research and development purposes. This publication is concerned with radiation safety issues and not with the uses of irradiators, nor does it cover the irradiation of product or its quality management. The five types of irradiator are: - Panoramic dry source storage irradiators; - Underwater irradiators, in which both the source and the product being irradiated are under water; - Panoramic wet source storage irradiators; - Electron beam irradiation facilities, in which irradiation is performed in an area that is potentially accessible to personnel, but that is kept inaccessible during the irradiation process; - X ray irradiation facilities, in which irradiation is performed in an area that is potentially accessible to personnel, but that is kept inaccessible during the irradiation process. Consideration of non-radiation-related risks and of the benefits resulting from the operation of irradiators is outside the scope of this Safety Guide. The practices of radiotherapy and radiography are also outside the scope of this Safety Guide. Category I gamma irradiators (i.e. 'self-shielded' irradiators) are outside the scope of this Safety Guide

  4. Heat removal tests on dry storage facilities for nuclear spent fuels

    International Nuclear Information System (INIS)

    Wataru, M.; Saegusa, T.; Koga, T.; Sakamoto, K.; Hattori, Y.

    1999-01-01

    In Japan, spent fuel generated in NPP is controlled and stored in dry storage facility away-from reactor. Natural convection cooling system of the storage facility is considered advantageous from both safety and economic point of view. In order to realize this type of facility it is necessary to develop an evaluation method for natural convection characteristics and to make a rational design taking account safety and economic factors. Heat removal tests with the reduces scale models of storage facilities (cask, vault and silo) identified the the flow pattern in the test modules. The temperature and velocity distributions were obtained and the heat transfer characteristics were evaluated

  5. Thermal-hydraulic tests with out-of-pile test facility for BOCA development

    International Nuclear Information System (INIS)

    Kitagishi, Shigeru; Aoyama, Masashi; Tobita, Masahiro; Inaba, Yoshitomo; Yamaura, Takayuki

    2012-01-01

    The fuel transient test facility was prepared for power ramping tests of light-water-reactor (LWR) fuels in the Japan Materials Testing Reactor (JMTR) under a contract project with the Nuclear Industrial Safety Agent (NISA) of the Ministry of Economy, Trade and Industry (METI). It is necessary to develop high accuracy analysis procedure for power ramping tests after restart of the JMTR. The out-of-pile test facility to simulate thermal-hydraulic conditions of the fuel transient test facility was therefore developed. Applicability of the analysis code ACE-3D was examined for thermal-hydraulic analysis of power ramping tests for 10x10 BWR fuels by the fuel transient test facility. As the results, the calculated temperature was 304°C in comparison with measured value of 304.9-317.4°C in the condition of 600 W/cm. There is a bright prospect of high accuracy power ramping tests by the fuel transient test facility in JMTR. (author)

  6. A test matrix sequencer for research test facility automation

    Science.gov (United States)

    Mccartney, Timothy P.; Emery, Edward F.

    1990-01-01

    The hardware and software configuration of a Test Matrix Sequencer, a general purpose test matrix profiler that was developed for research test facility automation at the NASA Lewis Research Center, is described. The system provides set points to controllers and contact closures to data systems during the course of a test. The Test Matrix Sequencer consists of a microprocessor controlled system which is operated from a personal computer. The software program, which is the main element of the overall system is interactive and menu driven with pop-up windows and help screens. Analog and digital input/output channels can be controlled from a personal computer using the software program. The Test Matrix Sequencer provides more efficient use of aeronautics test facilities by automating repetitive tasks that were once done manually.

  7. DOE standard: Filter test facility quality program plan

    International Nuclear Information System (INIS)

    1999-02-01

    This standard was developed primarily for application in US Department of Energy programs. It contains specific direction for HEPA filter testing performed at a DOE-accepted HEPA Filter Test Facility (FTF). Beneficial comments (recommendations, additions, deletions) and any pertinent data that may improve this document should be sent to the Office of Nuclear Safety Policy and Standards (EH-31), US Department of Energy, Washington, DC 20585, by letter or by using the self-addressed Document Improvement Proposal form (DOE F 1300.3) appearing at the end of this document

  8. Performance of smokeless gasoline fire test facility

    International Nuclear Information System (INIS)

    Griffin, J.F.; Watkins, R.A.

    1978-01-01

    Packaging for radioactive materials must perform satisfactorily when subjected to temperatures simulating an accident involving a fire. The new thermal test facility has proved to be a reliable method for satisfactorily performing the required test. The flame provides sufficient heat to assure that the test is valid, and the temperature can be controlled satisfactorily. Also, the air and water mist systems virtually eliminate any smoke and thereby exceed the local EPA requirements. The combination of the two systems provides an inexpensive, low maintenance technique for elimination of the smoke plume

  9. Tritium Systems Test Facility. Volume I

    International Nuclear Information System (INIS)

    Anderson, G.W.; Battleson, K.W.; Bauer, W.

    1976-10-01

    Sandia Laboratories proposes to build and operate a Tritium Systems Test Facility (TSTF) in its newly completed Tritium Research Laboratory at Livermore, California (see frontispiece). The facility will demonstrate at a scale factor of 1:200 the tritium fuel cycle systems for an Experimental Power Reactor (EPR). This scale for each of the TSTF subsystems--torus, pumping system, fuel purifier, isotope separator, and tritium store--will allow confident extrapolation to EPR dimensions. Coolant loop and reactor hall cleanup facilities are also reproduced, but to different scales. It is believed that all critical details of an EPR tritium system will be simulated correctly in the facility. Tritium systems necessary for interim devices such as the Ignition Test Reactor (ITR) or The Next Step (TNS) can also be simulated in TSTF at other scale values. The active tritium system will be completely enclosed in an inert atmosphere glove box which will be connected to the existing Gas Purification System (GPS) of the Tritium Research Laboratory. In effect, the GPS will become the scaled environmental control system which otherwise would have to be built especially for the TSTF

  10. Kauai Test Facility hazards assessment document

    International Nuclear Information System (INIS)

    Swihart, A.

    1995-05-01

    The Department of Energy Order 55003A requires facility-specific hazards assessment be prepared, maintained, and used for emergency planning purposes. This hazards assessment document describes the chemical and radiological hazards associated with the Kauai Test Facility, Barking Sands, Kauai, Hawaii. The Kauai Test Facility's chemical and radiological inventories were screened according to potential airborne impact to onsite and offsite individuals. The air dispersion model, ALOHA, estimated pollutant concentrations downwind from the source of a release, taking into consideration the toxicological and physical characteristics of the release site, the atmospheric conditions, and the circumstances of the release. The greatest distance to the Early Severe Health Effects threshold is 4.2 kilometers. The highest emergency classification is a General Emergency at the open-quotes Main Complexclose quotes and a Site Area Emergency at the Kokole Point Launch Site. The Emergency Planning Zone for the open-quotes Main Complexclose quotes is 5 kilometers. The Emergency Planning Zone for the Kokole Point Launch Site is the Pacific Missile Range Facility's site boundary

  11. 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

  12. Waste Sampling and Characterization Facility (WSCF) Complex Safety Analysis

    International Nuclear Information System (INIS)

    MELOY, R.T.

    2003-01-01

    The Waste Sampling and Characterization Facility (WSCF) is an analytical laboratory complex on the Hanford Site that was constructed to perform chemical and low-level radiological analyses on a variety of sample media in support of Hanford Site customer needs. The complex is located in the 600 area of the Hanford Site, east of the 200 West Area. Customers include effluent treatment facilities, waste disposal and storage facilities, and remediation projects. Customers primarily need analysis results for process control and to comply with federal, Washington State, and US. Department of Energy (DOE) environmental or industrial hygiene requirements. This document was prepared to analyze the facility for safety consequences and includes the following steps: Determine radionuclide and highly hazardous chemical inventories; Compare these inventories to the appropriate regulatory limits; Document the compliance status with respect to these limits; and Identify the administrative controls necessary to maintain this status

  13. Ensuring the safety of nuclear facilities located in large cities

    International Nuclear Information System (INIS)

    Ryazantsev, E.P.; Kolyadin, V.I.; Bylkin, B.K.; Zverkov, Yu.A.

    2002-01-01

    The problems of ensuring the safety of nuclear facilities and other facilities representing a radiation hazard (hereinafter referred to as 'nuclear facilities') which are located in large cities are considered in the light of the experience with the 'Kurchatov Institute' Russian Research Centre. The accumulation of substantial quantities of spent nuclear fuel and radwaste at the Centre was an inevitable consequence of the military and civilian nuclear research programmes which started there in 1943. A comprehensive programme has been developed for reducing the impact of ionizing radiation on the Centre's personnel, the population living near the Centre and the local environment. The authors describe the basic elements of a programme for decommissioning reactor facilities and eliminating spent fuel and radwaste storage sites and also describe how the programme is progressing. (author)

  14. Safety in acoustic emission testing

    International Nuclear Information System (INIS)

    Pollock, A.A.

    2004-01-01

    The human cost of accidents - the loss of family members and friends, and the impairment of life through injuries - makes prevention a very high priority in our society. Especially in the more industrialized countries, where high levels of personal comfort and security are the norm, resources are available to develop safety-enhancing technologies, cultures and management techniques. Thus, safety programs have become a well-established part of the industrial workplace. (author)

  15. E-ELT M1 test facility

    Science.gov (United States)

    Dimmler, M.; Marrero, J.; Leveque, S.; Barriga, P.; Sedghi, B.; Mueller, M.

    2012-09-01

    During the advanced design phase of the European Extremely Large Telescope (E-ELT) several critical components have been prototyped. During the last year some of them have been tested in dedicated test stands. In particular, a representative section of the E-ELT primary mirror has been assembled with 2 active and 2 passive segments. This test stand is equipped with complete prototype segment subunits, i.e. including support mechanisms, glass segments, edge sensors, position actuators as well as additional metrology for monitoring. The purpose is to test various procedures such as calibration, alignment and handling and to study control strategies. In addition the achievable component and subsystem performances are evaluated, and interface issues are identified. In this paper an overview of the activities related to the E-ELT M1 Test Facility will be given. Experiences and test results are presented.

  16. Large-coil-test-facility fault-tree analysis

    International Nuclear Information System (INIS)

    1982-01-01

    An operating-safety study is being conducted for the Large Coil Test Facility (LCTF). The purpose of this study is to provide the facility operators and users with added insight into potential problem areas that could affect the safety of personnel or the availability of equipment. This is a preliminary report, on Phase I of that study. A central feature of the study is the incorporation of engineering judgements (by LCTF personnel) into an outside, overall view of the facility. The LCTF was analyzed in terms of 32 subsystems, each of which are subject to failure from any of 15 generic failure initiators. The study identified approximately 40 primary areas of concern which were subjected to a computer analysis as an aid in understanding the complex subsystem interactions that can occur within the facility. The study did not analyze in detail the internal structure of the subsystems at the individual component level. A companion study using traditional fault tree techniques did analyze approximately 20% of the LCTF at the component level. A comparison between these two analysis techniques is included in Section 7

  17. Enhancement of safety at nuclear facilities in Pakistan

    International Nuclear Information System (INIS)

    Ahmad, S.A.; Hayat, T.; Azhar, W.

    2006-01-01

    Pakistan is benefiting from nuclear technology mostly in health and energy sectors as well as agriculture and industry and has an impeccable safety record. At the national level uses of nuclear technology started in 1955 resulting in the operation of Karachi Radioisotope Center, Karachi, in December 1960. Pakistan Nuclear Safety Committee (PNSC) was formulated in 1964 with subsequent promulgation of Pakistan Atomic Energy Commission (PAEC) Ordinance in 1965 to cope with the anticipated introduction of a research reactor, namely PARR-I, and a nuclear power plant, namely KANUPP. Since then Pakistan's nuclear program has expanded to include numerous nuclear facilities of varied nature. This program has definite economic and social impacts by producing electricity, treating and diagnosing cancer patients, and introducing better crop varieties. Appropriate radiation protection includes a number of measures including database of sealed radiation sources at PAEC operated nuclear facilities, see Table l, updated during periodic physical verification of these sources, strict adherence to the BSS-115, IAEA recommended enforcement of zoning at research reactors and NPPs, etc. Pakistan is party to several international conventions and treaties, such as Convention of Nuclear Safety and Early Notification, to improve and enhance safety at its nuclear facilities. In addition Pakistan generally and PAEC particularly believes in a blend of prudent regulations and good/best practices. This is described in this paper. (Author)

  18. Safety of Long-term Interim Storage Facilities - Workshop Proceedings

    International Nuclear Information System (INIS)

    2014-01-01

    The objective of this workshop was to discuss and review current national activities, plans and regulatory approaches for the safety of long term interim storage facilities dedicated to spent nuclear fuel (SF), high level waste (HLW) and other radioactive materials with prolonged storage regimes. It was also intended to discuss results of experiments and to identify necessary R and D to confirm safety of fuel and cask during the long-term storage. Safety authorities and their Technical Support Organisation (TSO), Fuel Cycle Facilities (FCF) operating organisations and international organisations were invited to share information on their approaches, practices and current developments. The workshop was organised in an opening session, three technical sessions, and a conclusion session. The technical sessions were focused on: - National approaches for long term interim storage facilities; - Safety requirements, regulatory framework and implementation issues; - Technical issues and operational experience, needs for R and D. Each session consisted of a number of presentations followed by a panel discussion moderated by the session Chairs. A summary of each session and subsequent discussion that ensued are provided as well as a summary of the results of the workshop with the text of the papers given and presentations made

  19. Comprehensive safety cases for radioactive waste management facilities

    International Nuclear Information System (INIS)

    Woollam, P.B.; Cameron, H.M.; Davies, A.R.; Hiscox, A.W.

    1995-01-01

    Probabilistic safety assessment methodology has been applied by Nuclear Electric plc (NE) to the development of comprehensive safety cases for the radioactive waste management processing and accumulation facilities associated with its 26 reactor systems. This paper describes the methodology and the safety case assessment criteria employed by NE. An overview of the results is presented, together with more detail of a specific safety analysis: storage of fuel element debris. No risk to the public greater than 10 -6 /y has been identified and the more significant risks arise from the potential for radioactive waste fires. There are no unacceptable risks from external hazards such as flooding, aircrash or seismic events. Some operations previously expected to have significant risks in fact have negligible risks, while the few faults with risks exceeding the assessment criteria were only identified as a result of this study

  20. Safety issues relating to the design of fusion power facilities

    International Nuclear Information System (INIS)

    Stasko, R.R.; Wong, K.Y.; Russell, S.B.

    1986-06-01

    In order to make fusion power a viable future source of energy, it will be necessary to ensure that the cost of power for fusion electric generation is competitive with advanced fission concepts. In addition, fusion power will have to live up to its original promise of being a more radiologically benign technology than fission, and be able to demonstrate excellent operational safety performance. These two requirements are interrelated, since the selection of an appropriate safety philosophy early in the design phase could greatly reduce or eliminate the capital costs of elaborate safety related and protective sytems. This paper will briefly overview a few of the key safety issues presently recognized as critical to the ultimate achievement of licensable, environmentally safe and socially acceptable fusion power facilities. 12 refs

  1. Documents pertaining to safety control of nuclear facilities

    International Nuclear Information System (INIS)

    1998-01-01

    The Finnish Radiation and Nuclear Safety Authority (STUK) controls the safety of nuclear facilities in Finland. This control encompasses on one hand the evaluation of plant safety on the basis of plans and analyses pertaining to the plant and on the other hand the inspection of plant structures, systems and components as well as of operational activity. STUK also monitors plants operational experience feedback and technical developments in the field, as well as the development of safety research and takes the necessary measures on their basis. Guide YVL 1.1 describes how STUK controls the design, construction and operation of nuclear power plants. The documents to be submitted to STUK are described in the nuclear energy legislation and YVL guides. This guide presents the mode of delivery, quality, contents and number of documents to be submitted to STUK

  2. Safety assessment of a vault-based disposal facility using the ISAM methodology

    International Nuclear Information System (INIS)

    Kelly, E.; Kim, C.-L.; Lietava, P.; Little, R.; Simon, I.

    2002-01-01

    As part of the IAEA's Co-ordinated Research Project (CRP) on Improving Long-term of Safety Assessment Methodologies for Near Surface Waste Disposal Facilities (ISAM), three example cases were developed. The aim was to testing the ISAM safety assessment methodology using as realistic as possible data. One of the Test Cases, the Vault Test Case (VTC), related to the disposal of low level radioactive waste (LLW) to a hypothetical facility comprising a set of above surface vaults. This paper uses the various steps of the ISAM safety assessment methodology to describe the work undertaken by ISAM participants in developing the VTC and provides some general conclusions that can be drawn from the findings of their work. (author)

  3. Evaluation of cold testing for Tokai Vitrification Facility

    International Nuclear Information System (INIS)

    Yoshioka, Masahiro; Inada, Eiichi

    1994-01-01

    The cold testing of the Tokai Vitrification Facility (TVF) was completed at the end of March, 1994 through the tests of nearly two years since May in 1992. The cold testing was carried out in order to evaluate the process equipment, product quality control, remote maintenance capability. The test results shown that TVF has enough performance with safety to treat the liquid waste in each process, and to control the product quality. For the remote maintenance of process equipment in the vitrification cell, the remote maintenance capability was confirmed for all remote equipment in the cell. The improvements were taken for some equipment with problem from the point of the operability and maintenance. It was confirmed by these test results that the TVF can go forward to the hot test operation using actual waste. (author)

  4. Technical Safety Requirements for the Waste Storage Facilities

    International Nuclear Information System (INIS)

    Laycak, D.T.

    2010-01-01

    This document contains Technical Safety Requirements (TSR) for the Radioactive and Hazardous Waste Management (RHWM) WASTE STORAGE FACILITIES, which include Area 625 (A625) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area at Lawrence Livermore National Laboratory (LLNL). The TSRs constitute requirements regarding the safe operation of the WASTE STORAGE FACILITIES. These TSRs are derived from the Documented Safety Analysis for the Waste Storage Facilities (DSA) (LLNL 2009). The analysis presented therein determined that the WASTE STORAGE FACILITIES are low-chemical hazard, Hazard Category 2 non-reactor nuclear facilities. The TSRs consist primarily of inventory limits and controls to preserve the underlying assumptions in the hazard and accident analyses. Further, appropriate commitments to safety programs are presented in the administrative controls sections of the TSRs. The WASTE STORAGE FACILITIES are used by RHWM to handle and store hazardous waste, TRANSURANIC (TRU) WASTE, LOW-LEVEL WASTE (LLW), mixed waste, California combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL as well as small amounts from other U.S. Department of Energy (DOE) facilities, as described in the DSA. In addition, several minor treatments (e.g., size reduction and decontamination) are carried out in these facilities. The WASTE STORAGE FACILITIES are located in two portions of the LLNL main site. A625 is located in the southeast quadrant of LLNL. The A625 fenceline is approximately 225 m west of Greenville Road. The DWTF Storage Area, which includes Building 693 (B693), Building 696 Radioactive Waste Storage Area (B696R), and associated yard areas and storage areas within the yard, is located in the northeast quadrant of LLNL in the DWTF complex. The DWTF Storage Area fenceline is approximately 90 m west of Greenville Road. A625 and the DWTF Storage Area are subdivided into various facilities and storage areas, consisting

  5. Technical Safety Requirements for the Waste Storage Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Laycak, D T

    2008-06-16

    This document contains Technical Safety Requirements (TSR) for the Radioactive and Hazardous Waste Management (RHWM) WASTE STORAGE FACILITIES, which include Area 625 (A625) and the Decontamination and Waste Treatment Facility (DWTF) Storage Area at Lawrence Livermore National Laboratory (LLNL). The TSRs constitute requirements regarding the safe operation of the WASTE STORAGE FACILITIES. These TSRs are derived from the 'Documented Safety Analysis for the Waste Storage Facilities' (DSA) (LLNL 2008). The analysis presented therein determined that the WASTE STORAGE FACILITIES are low-chemical hazard, Hazard Category 2 non-reactor nuclear facilities. The TSRs consist primarily of inventory limits and controls to preserve the underlying assumptions in the hazard and accident analyses. Further, appropriate commitments to safety programs are presented in the administrative controls sections of the TSRs. The WASTE STORAGE FACILITIES are used by RHWM to handle and store hazardous waste, TRANSURANIC (TRU) WASTE, LOW-LEVEL WASTE (LLW), mixed waste, California combined waste, nonhazardous industrial waste, and conditionally accepted waste generated at LLNL as well as small amounts from other U.S. Department of Energy (DOE) facilities, as described in the DSA. In addition, several minor treatments (e.g., size reduction and decontamination) are carried out in these facilities. The WASTE STORAGE FACILITIES are located in two portions of the LLNL main site. A625 is located in the southeast quadrant of LLNL. The A625 fenceline is approximately 225 m west of Greenville Road. The DWTF Storage Area, which includes Building 693 (B693), Building 696 Radioactive Waste Storage Area (B696R), and associated yard areas and storage areas within the yard, is located in the northeast quadrant of LLNL in the DWTF complex. The DWTF Storage Area fenceline is approximately 90 m west of Greenville Road. A625 and the DWTF Storage Area are subdivided into various facilities and storage areas

  6. The Brookhaven National Laboratory Accelerator Test Facility

    International Nuclear Information System (INIS)

    Batchelor, K.

    1992-01-01

    The Brookhaven National Laboratory Accelerator Test Facility comprises a 50 MeV traveling wave electron linear accelerator utilizing a high gradient, photo-excited, raidofrequency electron gun as an injector and an experimental area for study of new acceleration methods or advanced radiation sources using free electron lasers. Early operation of the linear accelerator system including calculated and measured beam parameters are presented together with the experimental program for accelerator physics and free electron laser studies

  7. Underground large scale test facility for rocks

    International Nuclear Information System (INIS)

    Sundaram, P.N.

    1981-01-01

    This brief note discusses two advantages of locating the facility for testing rock specimens of large dimensions in an underground space. Such an environment can be made to contribute part of the enormous axial load and stiffness requirements needed to get complete stress-strain behavior. The high pressure vessel may also be located below the floor level since the lateral confinement afforded by the rock mass may help to reduce the thickness of the vessel

  8. 40 CFR 792.43 - Test system care facilities.

    Science.gov (United States)

    2010-07-01

    .... (a) A testing facility shall have a sufficient number of animal rooms or other test system areas, as... different tests. (b) A testing facility shall have a number of animal rooms or other test system areas... waste and refuse or for safe sanitary storage of waste before removal from the testing facility...

  9. 305 Building Cold Test Facility Management Plan

    International Nuclear Information System (INIS)

    Whitehurst, R.

    1994-01-01

    This document provides direction for the conduct of business in Building 305 for cold testing tools and equipment. The Cold Test Facility represents a small portion of the overall building, and as such, the work instructions already implemented in the 305 Building will be utilized. Specific to the Cold Test there are three phases for the tools and equipment as follows: 1. Development and feature tests of sludge/fuel characterization equipment, fuel containerization equipment, and sludge containerization equipment to be used in K-Basin. 2. Functional and acceptance tests of all like equipment to be installed and operated in K-Basin. 3. Training and qualification of K-Basin Operators on equipment to be installed and operated in the Basin

  10. Clemson University Wind Turbine Drivetrain Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Tuten, James Maner [Clemson Univ., SC (United States); Haque, Imtiaz [Clemson Univ., SC (United States); Rigas, Nikolaos [Clemson Univ., SC (United States)

    2016-03-30

    In November of 2009, Clemson University was awarded a competitive grant from the U.S. Department of Energy to design, build and operate a facility for full-scale, highly accelerated mechanical testing of next-generation wind turbine drivetrain technologies. The primary goal of the project was to design, construct, commission, and operate a state-of-the-art sustainable facility that permits full-scale highly accelerated testing of advanced drivetrain systems for large wind turbines. The secondary goal was to meet the objectives of the American Recovery and Reinvestment Act of 2009, especially in job creation, and provide a positive impact on economically distressed areas in the United States, and preservation and economic recovery in an expeditious manner. The project was executed according to a managed cooperative agreement with the Department of Energy and was an extraordinary success. The resultant new facility is located in North Charleston, SC, providing easy transportation access by rail, road or ship and operates on an open access model such that it is available to the U.S. Wind Industry for research, analysis, and evaluation activities. The 72 m by 97 m facility features two mechanical dynamometer test bays for evaluating the torque and blade dynamic forces experienced by the rotors of wind turbine drivetrains. The dynamometers are rated at 7.5 MW and 15 MW of low speed shaft power and are configured as independent test areas capable of simultaneous operation. All six degrees of freedom, three linear and three rotational, for blade and rotor dynamics are replicated through the combination of a drive motor, speed reduction gearbox and a controllable hydraulic load application unit (LAU). This new LAU setup readily supports accelerated lifetime mechanical testing and load analysis for the entire drivetrain system of the nacelle and easily simulates a wide variety of realistic operating scenarios in a controlled laboratory environment. The development of these

  11. Simulation Facilities and Test Beds for Galileo

    Science.gov (United States)

    Schlarmann, Bernhard Kl.; Leonard, Arian

    2002-01-01

    Galileo is the European satellite navigation system, financed by the European Space Agency (ESA) and the European Commission (EC). The Galileo System, currently under definition phase, will offer seamless global coverage, providing state-of-the-art positioning and timing services. Galileo services will include a standard service targeted at mass market users, an augmented integrity service, providing integrity warnings when fault occur and Public Regulated Services (ensuring a continuity of service for the public users). Other services are under consideration (SAR and integrated communications). Galileo will be interoperable with GPS, and will be complemented by local elements that will enhance the services for specific local users. In the frame of the Galileo definition phase, several system design and simulation facilities and test beds have been defined and developed for the coming phases of the project, respectively they are currently under development. These are mainly the following tools: Galileo Mission Analysis Simulator to design the Space Segment, especially to support constellation design, deployment and replacement. Galileo Service Volume Simulator to analyse the global performance requirements based on a coverage analysis for different service levels and degrades modes. Galileo System Simulation Facility is a sophisticated end-to-end simulation tool to assess the navigation performances for a complete variety of users under different operating conditions and different modes. Galileo Signal Validation Facility to evaluate signal and message structures for Galileo. Galileo System Test Bed (Version 1) to assess and refine the Orbit Determination &Time Synchronisation and Integrity algorithms, through experiments relying on GPS space infrastructure. This paper presents an overview on the so called "G-Facilities" and describes the use of the different system design tools during the project life cycle in order to design the system with respect to

  12. A proactive method for safety management in nuclear facilities

    International Nuclear Information System (INIS)

    Grecco, Claudio Henrique dos Santos; Carvalho, Paulo Victor Rodrigues de; Santos, Isaac Antonio Luquetti dos

    2014-01-01

    Due to the modern approach to address the safety of nuclear facilities which highlights that these organizations must be able to assess and proactively manage their activities becomes increasingly important the need for instruments to evaluate working conditions. In this context, this work presents a proactive method of managing organizational safety, which has three innovative features: 1) the use of predictive indicators that provide current information on the performance of activities, allowing preventive actions and not just reactive in safety management, different from safety indicators traditionally used (reactive indicators) that are obtained after the occurrence of undesired events; 2) the adoption of resilience engineering approach in the development of indicators - indicators are based on six principles of resilience engineering: top management commitment, learning, flexibility, awareness, culture of justice and preparation for the problems; 3) the adoption of the concepts and properties of fuzzy set theory to deal with subjectivity and consistency of human trials in the evaluation of the indicators. The fuzzy theory is used primarily to map qualitative models of decision-making, and inaccurate representation methods. The results of this study aim an improvement in performance and safety in organizations. The method was applied in a radiopharmaceutical shipping sector of a nuclear facility. The results showed that the method is a good monitoring tool objectively and proactively of the working conditions of an organizational domain

  13. Radiation safety program in a high dose rate brachytherapy facility

    International Nuclear Information System (INIS)

    Rodriguez, L.V.; Hermoso, T.M.; Solis, R.C.

    2001-01-01

    The use of remote afterloading equipment has been developed to improve radiation safety in the delivery of treatment in brachytherapy. Several accidents, however, have been reported involving high dose-rate brachytherapy system. These events, together with the desire to address the concerns of radiation workers, and the anticipated adoption of the International Basic Safety Standards for Protection Against Ionizing Radiation (IAEA, 1996), led to the development of the radiation safety program at the Department of Radiotherapy, Jose R. Reyes Memorial Medical Center and at the Division of Radiation Oncology, St. Luke's Medical Center. The radiation safety program covers five major aspects: quality control/quality assurance, radiation monitoring, preventive maintenance, administrative measures and quality audit. Measures for evaluation of effectiveness of the program include decreased unnecessary exposures of patients and staff, improved accuracy in treatment delivery and increased department efficiency due to the development of staff vigilance and decreased anxiety. The success in the implementation required the participation and cooperation of all the personnel involved in the procedures and strong management support. This paper will discuss the radiation safety program for a high dose rate brachytherapy facility developed at these two institutes which may serve as a guideline for other hospitals intending to install a similar facility. (author)

  14. Fast Flux Test Facility primary sodium valves

    International Nuclear Information System (INIS)

    Rabe, G.B.; Ezra, B.C.

    1977-01-01

    The design and development of the valves used in the primary sodium coolant loop of the Fast Flux Test Facility is described. One tilting-disk check valve is used in the cold leg of the coolant loop. It is designed to limit flow reversal in the loop while maintaining a low pressure drop during forward flow. Two isolation valves are used in each coolant loop--one in the cold leg and one in the hot leg. They are of the motor-operated swinging-gate type. The design, analysis, and testing programs undertaken to develop and qualify these valves are described

  15. Alpha Fuels Environmental Test Facility impact gun

    International Nuclear Information System (INIS)

    Anderson, C.G.

    1978-01-01

    The Alpha Fuels Environmental Test Facility (AFETF) impact gun is a unique tool for impact testing 238 PuO 2 -fueled heat sources of up to 178-mm dia at velocities to 300 m/s. An environmentally-sealed vacuum chamber at the muzzle of the gun allows preheating of the projectile to 1,000 0 C. Immediately prior to impact, the heat source projectile is completely sealed in a vacuum-tight catching container to prevent escape of its radioactive contents should rupture occur. The impact velocity delivered by this gas-powered gun can be regulated to within +-2%

  16. Use of risk information to safety regulation. Fabrication facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    A procedure of ISA (Integrated Safety Analysis) for uranium fuel fabrication/enrichment facilities has been under the development aiming to utilize risk information for safety regulations in this project. Activities in the fiscal year 2012 are summarized in the paper. There are two major activities in the year. First one is a study on ISA procedure for external events such as earthquakes. Second one is that for chemical consequences such as UF6 and HF. Other than the activities a fundamental study on a policy of utilizing risk information was conducted. The outline and results are provided in the chapter 1 and 2 respectively. (author)

  17. The Management System for Facilities and Activities. Safety Requirements

    International Nuclear Information System (INIS)

    2011-01-01

    This publication establishes requirements for management systems that integrate safety, health, security, quality assurance and environmental objectives. A successful management system ensures that nuclear safety matters are not dealt with in isolation but are considered within the context of all these objectives. The aim of this publication is to assist Member States in establishing and implementing effective management systems that integrate all aspects of managing nuclear facilities and activities in a coherent manner. It details the planned and systematic actions necessary to provide adequate confidence that all these requirements are satisfied. Contents: 1. Introduction; 2. Management system; 3. Management responsibility; 4. Resource management; 5. Process implementation; 6. Measurement, assessment and improvement.

  18. Importance of the licensing process on the safety culture in the Brazilian nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    Motta, E.S.; Sousa, A.L.B. de; Paiva, R.L.C. de; Mezrahi, A.

    2013-01-01

    The main objective of the Nuclear Fuel Cycle Facilities licensing processes is to ensure the safety of these installations in their entire life cycle (in the installation site selection, designing, construction, pre-operational tests, operational and decommissioning phases). The Brazilian licensing process requires from the operator, among others, before the operating license: (I) a Site Report and a Final Safety Analysis Report, ensuring that all safety related issues are adequately analyzed and understood; (II) a formal structured Management System focused on the installation safety; and (III) dissemination of safety related information to all involved operator employees and subcontractors. Therefore, these requirements reflect in an adequate operator actions and practices, ensuring a working environment with a high level of safety culture. (author)

  19. Importance of the licensing process on the safety culture in the Brazilian nuclear fuel cycle facilities

    Energy Technology Data Exchange (ETDEWEB)

    Motta, E.S.; Sousa, A.L.B. de; Paiva, R.L.C. de; Mezrahi, A., E-mail: emotta@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    The main objective of the Nuclear Fuel Cycle Facilities licensing processes is to ensure the safety of these installations in their entire life cycle (in the installation site selection, designing, construction, pre-operational tests, operational and decommissioning phases). The Brazilian licensing process requires from the operator, among others, before the operating license: (I) a Site Report and a Final Safety Analysis Report, ensuring that all safety related issues are adequately analyzed and understood; (II) a formal structured Management System focused on the installation safety; and (III) dissemination of safety related information to all involved operator employees and subcontractors. Therefore, these requirements reflect in an adequate operator actions and practices, ensuring a working environment with a high level of safety culture. (author)

  20. Test for radioactive material transport package safety

    International Nuclear Information System (INIS)

    Li Guoqiang; Zhao Bing; Zhang Jiangang; Wang Xuexin; Ma Anping

    2012-01-01

    Regulations on radioactive material transport in China were introduced. Test facilities and data acquiring instruments for radioactive material package in China Institute for Radiation Protection were also introduced in this paper, which were used in drop test and thermal test. Test facilities were constructed according to the requirements of IAEA's 'Regulations for the Safe Transport of Radioactive Material' (TS-R-l) and Chinese 'Regulations for the Safe Transport of Radioactive Material' (GB 11806-2004). Drop test facilities were used in free drop test, penetration test, mechanical test (free drop test Ⅰ, free drop test Ⅱ and free drop test Ⅲ) of type A and type B packages weighing less than thirteen tons. Thermal test of type B packages can be carried out in the thermal test facilities. Certification tests of type FCo70-YQ package, type 30A-HB-01 package, type SY-I package and type XAYT-I package according to regulations were done using these facilities. (authors)

  1. Radiation safety management system in a radioactive facility

    International Nuclear Information System (INIS)

    Amador, Zayda H.

    2008-01-01

    Full text: This paper illustrates the Cuban experience in implementing and promoting an effective radiation safety system for the Centre of Isotopes, the biggest radioactive facility of our country. Current management practice demands that an organization inculcate culture of safety in preventing radiation hazard. The aforementioned objectives of radiation protection can only be met when it is implemented and evaluated continuously. Commitment from the workforce to treat safety as a priority and the ability to turn a requirement into a practical language is also important to implement radiation safety policy efficiently. Maintaining and improving safety culture is a continuous process. There is a need to establish a program to measure, review and audit health and safety performance against predetermined standards. All those areas of the radiation protection program are considered (e.g. licensing and training of the staff, occupational exposure, authorization of the practices, control of the radioactive material, radiological occurrences, monitoring equipment, radioactive waste management, public exposure due to airborne effluents, audits and safety costs). A set of indicators designed to monitor key aspects of operational safety performance are used. Their trends over a period of time are analyzed with the modern information technologies, because this can provide an early warning to plant management for searching causes behind the observed changes. In addition to analyze the changes and trends, these indicators are compared against identified targets and goals to evaluate performance strengths and weaknesses. A structured and proper radiation self-auditing system is seen as a basic requirement to meet the current and future needs in sustainability of radiation safety. The integrated safety management system establishment has been identified as a goal and way for the continuous improvement. (author)

  2. Safety Research Experiment Facility Project. Conceptual design report. Volume II. Building and facilities

    International Nuclear Information System (INIS)

    1975-12-01

    The conceptual design of Safety Research Experiment Facility (SAREF) site system includes a review and evaluation of previous geotechnical reports for the area where SAREF will be constructed and the conceptual design of access and in-plant roads, parking, experiment-transport-vehicle maneuvering areas, security fencing, drainage, borrow area development and restoration, and landscaping

  3. 40 CFR 160.43 - Test system care facilities.

    Science.gov (United States)

    2010-07-01

    ... testing facility shall have a number of animal rooms or other test system areas separate from those... housed, facilities shall exist for the collection and disposal of all animal waste and refuse or for safe sanitary storage of waste before removal from the testing facility. Disposal facilities shall be so...

  4. Periodic Safety Review in Interim Storage Facilities - Current Regulation and Experiences in Germany

    International Nuclear Information System (INIS)

    Neles, Julia Mareike; Schmidt, Gerhard

    2014-01-01

    Periodic safety reviews in nuclear power plants in Germany have been performed since the end of the 1980's as an indirect follow-up of the accident in Chernobyl and, in the meantime, are formally required by law. During this process the guidelines governing this review were developed in stages and reached their final form in 1996. Interim storage facilities and other nuclear facilities at that time were not included, so the guidelines were solely focused on the specific safety issues of nuclear power plants. Following IAEA's recommendations, the Western European Nuclear Regulator Association (WENRA) introduced PSRs in its safety reference levels for storage facilities (current version in WGWD report 2.1 as of Feb 2011: SRLs 59 - 61). Based on these formulations, Germany improved its regulation in 2010 with a recommendation of the Nuclear Waste Management Commission (Entsorgungskommission, ESK), an expert advisory commission for the federal regulatory body BMU. The ESK formulated these detailed requirements in the 'ESK recommendation for guides to the performance of periodic safety reviews for interim storage facilities for irradiated fuel elements and heat-generating radioactive waste'. Before finalization of the guideline a test phase was introduced, aimed to test the new regulation in practice and to later include the lessons learned in the final formulation of the guideline. The two-year test phase started in October 2011 in which the performance of a PSR will be tested at two selected interim storage facilities. Currently these recommendations are discussed with interested/concerned institutions. The results of the test phase shall be considered for improvements of the draft and during the final preparation of guidelines. Currently the PSR for the first ISF is in an advanced stage, the second facility just started the process. Preliminary conclusions from the test phase show that the implementation of the draft guideline requires interpretation. The aim of a

  5. Vitrification Facility integrated system performance testing report

    International Nuclear Information System (INIS)

    Elliott, D.

    1997-01-01

    This report provides a summary of component and system performance testing associated with the Vitrification Facility (VF) following construction turnover. The VF at the West Valley Demonstration Project (WVDP) was designed to convert stored radioactive waste into a stable glass form for eventual disposal in a federal repository. Following an initial Functional and Checkout Testing of Systems (FACTS) Program and subsequent conversion of test stand equipment into the final VF, a testing program was executed to demonstrate successful performance of the components, subsystems, and systems that make up the vitrification process. Systems were started up and brought on line as construction was completed, until integrated system operation could be demonstrated to produce borosilicate glass using nonradioactive waste simulant. Integrated system testing and operation culminated with a successful Operational Readiness Review (ORR) and Department of Energy (DOE) approval to initiate vitrification of high-level waste (HLW) on June 19, 1996. Performance and integrated operational test runs conducted during the test program provided a means for critical examination, observation, and evaluation of the vitrification system. Test data taken for each Test Instruction Procedure (TIP) was used to evaluate component performance against system design and acceptance criteria, while test observations were used to correct, modify, or improve system operation. This process was critical in establishing operating conditions for the entire vitrification process

  6. Preclosure radiological safety analysis for the exploratory shaft facilities

    International Nuclear Information System (INIS)

    Ma, C.W.; Miller, D.D.; Jardine, L.J.

    1992-06-01

    This study assesses which structures, systems, and components of the exploratory shaft facility (ESF) are important to safety when the ESF is converted to become part of the operating waste repository. The assessment follows the methodology required by DOE Procedure AP-6.10Q. Failures of the converted ESF during the preclosure period have been evaluated, along with other underground accidents, to determine the potential offsite radiation doses and associated probabilities. The assessment indicates that failures of the ESF will not result in radiation doses greater than 0.5 rem at the nearest unrestricted area boundary. Furthermore, credible accidents in other underground facilities will not result in radiation doses larger than 0.5 rem, even if any structure, system, or component of the converted ESF fails at the same time. Therefore, no structure, system, or component of the converted ESF is important to safety

  7. 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.

  8. 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.

  9. Digital tape unit test facility software

    Science.gov (United States)

    Jackson, J. T.

    1971-01-01

    Two computer programs are described which are used for the collection and analysis of data from the digital tape unit test facility (DTUTF). The data are the recorded results of skew tests made on magnetic digital tapes which are used on computers as input/output media. The results of each tape test are keypunched onto an 80 column computer card. The format of the card is checked and the card image is stored on a master summary tape via the DTUTF card checking and tape updating system. The master summary tape containing the results of all the tape tests is then used for analysis as input to the DTUTF histogram generating system which produces a histogram of skew vs. date for selected data, followed by some statistical analysis of the data.

  10. 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.

  11. Radiological safety assessment of a reference INTOR facility

    International Nuclear Information System (INIS)

    Khan, T.A.; Stasko, R.R.; Watts, R.T.; Shaw, G.; Morrison, C.A.; Russell, S.; Kempe, T.; Zimmerman, R.

    1985-03-01

    This report consists of a number of separate studies all of which were performed in support of INTOR Critical Issue D: Tritium Containment and Personnel Access vs Remote Maintenance. The common thread running through these studies is the radiological safety element in the design and operation of the INTOR facility. The intent is to help establish a firm basis for comparisons between a reactor cell maintenance option which requires personnel access, and one which involves completely remote maintenance

  12. In-pile experimental facility needs for LMFR safety research

    International Nuclear Information System (INIS)

    Kawata, Norio; Niwa, Hajime

    1994-01-01

    Although the achievement of the safety research during the past years has been significant, there still exists a strong need for future research, especially when there is prospect for future LMFR commercialization. In this paper, our current views are described on future research needs especially with a new in-pile experimental facility. The basic ideas and progress are outlined of a preliminary feasibility study. (author)

  13. The Great Plains Wind Power Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John [Texas Tech Univ., Lubbock, TX (United States)

    2014-01-30

    This multi-year, multi-faceted project was focused on the continued development of a nationally-recognized facility for the testing, characterization, and improvement of grid-connected wind turbines, integrated wind-water desalination systems, and related educational and outreach topics. The project involved numerous faculty and graduate students from various engineering departments, as well as others from the departments of Geosciences (in particular the Atmospheric Science Group) and Economics. It was organized through the National Wind Institute (NWI), which serves as an intellectual hub for interdisciplinary and transdisciplinary research, commercialization and education related to wind science, wind energy, wind engineering and wind hazard mitigation at Texas Tech University (TTU). Largely executed by an academic based team, the project resulted in approximately 38 peer-reviewed publications, 99 conference presentations, the development/expansion of several experimental facilities, and two provisional patents.

  14. Cryogenics for a vertical test stand facility for testing superconducting radio frequency cavities at RRCAT

    International Nuclear Information System (INIS)

    Gupta, Prabhat Kumar; Kumar, Manoj; Kush, P.K.

    2015-01-01

    Vertical Test Stand (VTS) Facility is located in a newly constructed building of Cryo-Engineering and Cryo-Module Development Division (CCDD). This test facility is one of the important facilities to develop SCRF technologies for superconducting accelerators like Indian Spallation Neutron Source. VTS has to be used for regular testing of the Superconducting Radio Frequency (SRF) Niobium cavities at nominal frequency of 1.3 GHz/ 650 MHz at 4 K / 2 K liquid helium (LHe) bath temperatures. Testing of these cavities at 2 K evaluates cavity processing methods, procedures and would also serve as a pre-qualification test for cavity to test it in horizontal cryostat, called horizontal test stand, with other cavity components such as tuner and helium vessel. Cryogenic technologies play a major role in these cavity testing facilities. Achieving and maintaining a stable temperature of 2 K in these test stands on regular and reliable basis is a challenging task and require broad range of cryogenic expertise, large scale system level understanding and many in-house technological and process developments. Furthermore this test stand will handle large amount of liquid helium. Therefore, an appropriately designed infrastructure is required to handle such large amount of helium gas generated during the operation of VTS .This paper describes the different cryogenic design aspects, initial cryogenic operation results and different cryogenic safety aspects. (author)

  15. Safety and radiation protection in mining and milling facilities

    Energy Technology Data Exchange (ETDEWEB)

    Magalhaes, Maisa H.; Schenato, Flavia; Cruz, Paulo R., E-mail: maisahm@cnen.gov.br, E-mail: schenato@cnen.gov.br, E-mail: pcruz@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil); Xavier, Ana M., E-mail: axavier@cnen.gov.br [Comissao Nacional de Energia Nuclear (ESPOA/CNEN-RS), Porto Alegre, RS (Brazil). Escritorio de Porto Alegre

    2011-07-01

    Federal Legislation in Brazil establishes that the Brazilian Nuclear Energy Commission - CNEN - is responsible for the surveillance of the industrialization of nuclear ores and the production and commerce of nuclear materials in such way that activities such as buying, selling, import and export, are subject to previous licensing and surveillance. Regulation CNEN-NN-4.01 on Safety and Radiation Protection in Mining and Milling Facilities of conventional ores containing naturally occurring radioactive materials, NORM, was issued in 2004 establishing both a methodology for classification of these facilities into three Categories, taking into account both the contents of uranium and thorium in the ores and the applicable radiation and safety requirements based on a graded approach. Although the lack of a licensing process in the above mentioned Regulation made its implementation a difficult task, CNEN, by means of an initial survey, identified ca. 30 mining and milling industries of conventional ores containing uranium and thorium with concentrations above 10 Bq/g. More recently, a new juridical understanding of the legislation concluded that CNEN must issue licences and authorizations for the possession and storage of all ores with uranium and thorium concentrations above exemption levels. A proper surveillance programme encompassing 13 of these mining facilities was then put forward aiming at the improvement of their safety and radiation protection. This article presents an overview of NORM exploitation in Brazil and put forward suggestions for achieving viable solutions for the protection of workers, general public and environment from the effects of ionizing radiation. (author)

  16. Safety analysis of the existing 804 and 845 firing facilities

    International Nuclear Information System (INIS)

    Odell, B.N.

    1986-01-01

    A safety analysis was performed to determine if normal operations and/or potential accidents at the 804 and 845 Firing Facilities at Site 300 could present undue hazards to the general public, peronnel at Site 300, or have an adverse effect on the environment. The normal operation and credible accident that might have an effect on these facilities or have off-site consequence were considered. It was determined by this analysis that all but one of the hazards were either low or of the type or magnitude routinely encountered and/or accepted by the public. The exception was explosives. Since this hazard has the potential for causing significant on-site and minimum off-site consequences, Bunkers 804 and 845 have been classified as moderate hazard facilties per DOE Order 5481.1A. This safety analysis concluded that the operation at these facilities will present no undue risk to the health and safety of LLNL employees or the public

  17. Results from the CLIC Test Facility

    CERN Document Server

    Braun, H; Bossart, Rudolf; Chautard, F; Corsini, R; Delahaye, J P; Godot, J C; Hutchins, S; Kamber, I; Madsen, J H B; Rinolfi, Louis; Rossat, G; Schreiber, S; Suberlucq, Guy; Thorndahl, L; Wilson, Ian H; Wuensch, Walter

    1996-01-01

    In order to study the principle of the Compact Linear Collider (CLIC) based on the Two Beam Acceleration (TBA) scheme at high frequency, a CLIC Test Facility (CTF) has been set-up at CERN. After four years of successful running, the experimental programme is now fully completed and all its objectives reached, particularly the generation of a high intensity drive beam with short bunches by a photo-injector, the production of 30 GHz RF power and the acceleration of a probe beam by 30 GHz structures. A summary of the CTF results and their impact on linear collider design is given. This covers 30 GHz high power testing, study of intense, short single bunches; as well as RF-Gun, photocathode and beam diagnostic developments. A second phase of the test facility (CTF2) is presently being installed to demonstrate the feasibility of the TBA scheme by constructing a fully engineered, 10 m long, test section very similar to the CLIC drive and main linacs, producing up to 480 MW of peak RF power at 30 GHz and acceleratin...

  18. Thermionic system evaluated test (TSET) facility description

    Science.gov (United States)

    Fairchild, Jerry F.; Koonmen, James P.; Thome, Frank V.

    1992-01-01

    A consortium of US agencies are involved in the Thermionic System Evaluation Test (TSET) which is being supported by the Strategic Defense Initiative Organization (SDIO). The project is a ground test of an unfueled Soviet TOPAZ-II in-core thermionic space reactor powered by electrical heat. It is part of the United States' national thermionic space nuclear power program. It will be tested in Albuquerque, New Mexico at the New Mexico Engineering Research Institute complex by the Phillips Laboratoty, Sandia National Laboratories, Los Alamos National Laboratory, and the University of New Mexico. One of TSET's many objectives is to demonstrate that the US can operate and test a complete space nuclear power system, in the electrical heater configuration, at a low cost. Great efforts have been made to help reduce facility costs during the first phase of this project. These costs include structural, mechanical, and electrical modifications to the existing facility as well as the installation of additional emergency systems to mitigate the effects of utility power losses and alkali metal fires.

  19. Developing a safety report for an existing conversion facility

    International Nuclear Information System (INIS)

    Carisse, Hess

    2013-01-01

    A review of the process used to meet the regulatory requirements for a Safety Report at an existing conversion facility is described. This paper will cover the establishment of the regulatory criteria, selection of appropriate methodologies, identification of events and modeling of credible events. Once established there is on-going maintenance to deal with design changes and the need for periodic reviews will also be discussed. Challenges in dealing with the various phases, including incorporation of historical licensing documents, and lessons learned are presented. Of specific interest is the failure of the selected methodology to deal with infrastructure issues. One aspect of lessons learned that will be explored is the lack of an available mechanism for sharing information with similar fuel cycle facilities which is compounded by the fact that there are a small number of fuel cycle facilities compared to nuclear power plants. Possible approaches to dealing with this issue are also discussed. (authors)

  20. Seismic safety of the LLL plutonium facility (Building 332)

    International Nuclear Information System (INIS)

    Torkarz, F.J.; Shaw, G.

    1980-01-01

    This report states the basis for the Lawrence Livermore Laboratory's assurance to the public that the plutonium operations at the Laboratory pose essentially no risk to anyone's health or safety, either under normal circumstances or in the event of an earthquake or a fire. The report is intended for a general audience, and so for the most part it is not highly technical. It summarizes the steps taken to ensure the seismic safety of the plutonium facility (Bldg. 332). It describes plutonium and its potential hazard and how the facility copes with that hazard. It recounts the geologic investigations and interpretations that led to the design-basis earthquake (DBE) for the Livermore site, and presents a summary analysis of the facility structure in relation to the DBE. An appendix presents a quantitative calculation of the health risk to the public associated with the worst-case hypothetical fire. The document supports the conclusions that the facility will continue to function safely after the maximum earthquake ground motion to which it may be subjected and that there is no evidence of a potential for surface offset under it

  1. Demonstration of safety of decommissioning of facilities using radioactive material

    International Nuclear Information System (INIS)

    Batandjieva, Borislava; O'Donnell, Patricio

    2008-01-01

    Full text:The development of nuclear industry worldwide in the recent years has particular impact on the approach of operators, regulators and interested parties to the implementation of the final phases (decommissioning) of all facilities that use radioactive material (from nuclear power plants, fuel fabrication facilities, research reactors to small research or medical laboratories). Decommissioning is becoming an increasingly important activity for two main reasons - termination of the practice in a safe manner with the view to use the facility or the site for other purposes, or termination of the practice and reuse the facility or site for new built nuclear facilities. The latter is of special relevance to multi-facility sites where for example new nuclear power plants and envisaged. However, limited countries have the adequate legal and regulatory framework, and experience necessary for decommissioning. In order to respond to this challenge of the nuclear industry and assist Member States in the adequate planning, conduct and termination of decommissioning of wide range of facilities, over the last decade the IAEA has implemented and initiated several projects in this field. One of the main focuses of this assistance to operators, regulators and specialists involved in decommissioning is the evaluation and demonstration of safety of decommissioning. This importance of these Agency activities was also highlighted in the International Action Plan on Decommissioning, during the second Joint Convention meeting in 2006 and the International Conference on Lessons Learned from Decommissioning in Athens in 2006. The IAEA has been providing technical support to its Member States in this field through several mechanisms: (1) the establishment of a framework of safety standards on decommissioning and development of a supporting technical documents; (2) the establishment of an international peer review mechanism for decommissioning; (3) the technical cooperation projects

  2. Facility for endurance tests of thermal insulations

    International Nuclear Information System (INIS)

    Mauersberger, R.

    1984-01-01

    In the following report the design and construction of an experimental facility for endurance tests of thermal insulations is presented. It's name in abbreviation is 'ADI' standing for the German words A nlage zum Dauertest von Isolierungen . This test facility was build by HRB in order to investigate the performance of thermal insulation systems of hot gas ducts for the process heat-reactor-project. The tests are intended to simulate the conditions of reactor operation. They include short-time experiments for selection of insulation-concepts and in a second step long-time experiments as performance tests. During these tests are measured the effective heat conductivity the local heat losses the temperature profiles of the insulation, of the fixing elements and along the wall of the duct. The design-data required to perform all these tasks are shown in the first picture: The gas-atmosphere must be Helium in tests like in reactor with regard to the special thermal and hydraulic properties of Helium and to the influence of Helium on mechanic friction and wear. The hot gas temperature in the PNP-reactor will be 950 deg. C and should be equal in the experiments. The temperature on the cold side of the insulation has to be adjustable from 50 deg. C up to 300 deg. C. The Helium pressure in the hot gas ducts of a HTR-plant is about 42 bar. The ADI was laid out for 70 bar to cover the hole range of interest. A Helium mass flow has to stream through the insulated test duct in order to realize equal temperatures on the hot side of the insulation. A flow rate of 4,5 kg/s is sufficient for this requirement. The axial pressure gradient along the insulation must be the same as in the reactor, because this has an essential influence on the heat losses. This pressure gradient is about 40 Pa/m

  3. Technical concept for a greater-confinement-disposal test facility

    International Nuclear Information System (INIS)

    Hunter, P.H.

    1982-01-01

    Greater confinement disposal (GCO) has been defined by the National Low-Level Waste Program as the disposal of low-level waste in such a manner as to provide greater containment of radiation, reduce potential for migration or dispersion or radionuclides, and provide greater protection from inadvertent human and biological intrusions in order to protect the public health and safety. This paper discusses: the need for GCD; definition of GCD; advantages and disadvantages of GCD; relative dose impacts of GCD versus shallow land disposal; types of waste compatible with GCD; objectives of GCD borehole demonstration test; engineering and technical issues; and factors affecting performance of the greater confinement disposal facility

  4. Operation of the nuclear fuel cycle test facilities -Operation of the hot test loop facilities

    International Nuclear Information System (INIS)

    Chun, S. Y.; Jeong, M. K.; Park, C. K.; Yang, S. K.; Won, S. Y.; Song, C. H.; Jeon, H. K.; Jeong, H. J.; Cho, S.; Min, K. H.; Jeong, J. H.

    1997-01-01

    A performance and reliability of a advanced nuclear fuel and reactor newly designed should be verified by performing the thermal hydraulics tests. In thermal hydraulics research team, the thermal hydraulics tests associated with the development of an advanced nuclear fuel and reactor haven been carried out with the test facilities, such as the Hot Test Loop operated under high temperature and pressure conditions, Cold Test Loop, RCS Loop and B and C Loop. The objective of this project is to obtain the available experimental data and to develop the advanced measuring techniques through taking full advantage of the facilities. The facilities operated by the thermal hydraulics research team have been maintained and repaired in order to carry out the thermal hydraulics tests necessary for providing the available data. The performance tests for the double grid type bottom end piece which was improved on the debris filtering effectivity were performed using the PWR-Hot Test Loop. The CANDU-Hot Test Loop was operated to carry out the pressure drop tests and strength tests of CANFLEX fuel. The Cold Test Loop was used to obtain the local velocity data in subchannel within HANARO fuel bundle and to study a thermal mixing characteristic of PWR fuel bundle. RCS thermal hydraulic loop was constructed and the experiments have been carried out to measure the critical heat flux. In B and C Loop, the performance tests for each component were carried out. (author). 19 tabs., 78 figs., 19 refs

  5. Operation of the nuclear fuel cycle test facilities -Operation of the hot test loop facilities

    Energy Technology Data Exchange (ETDEWEB)

    Chun, S. Y.; Jeong, M. K.; Park, C. K.; Yang, S. K.; Won, S. Y.; Song, C. H.; Jeon, H. K.; Jeong, H. J.; Cho, S.; Min, K. H.; Jeong, J. H.

    1997-01-01

    A performance and reliability of a advanced nuclear fuel and reactor newly designed should be verified by performing the thermal hydraulics tests. In thermal hydraulics research team, the thermal hydraulics tests associated with the development of an advanced nuclear fuel and reactor haven been carried out with the test facilities, such as the Hot Test Loop operated under high temperature and pressure conditions, Cold Test Loop, RCS Loop and B and C Loop. The objective of this project is to obtain the available experimental data and to develop the advanced measuring techniques through taking full advantage of the facilities. The facilities operated by the thermal hydraulics research team have been maintained and repaired in order to carry out the thermal hydraulics tests necessary for providing the available data. The performance tests for the double grid type bottom end piece which was improved on the debris filtering effectivity were performed using the PWR-Hot Test Loop. The CANDU-Hot Test Loop was operated to carry out the pressure drop tests and strength tests of CANFLEX fuel. The Cold Test Loop was used to obtain the local velocity data in subchannel within HANARO fuel bundle and to study a thermal mixing characteristic of PWR fuel bundle. RCS thermal hydraulic loop was constructed and the experiments have been carried out to measure the critical heat flux. In B and C Loop, the performance tests for each component were carried out. (author). 19 tabs., 78 figs., 19 refs.

  6. Reflooding phenomena of German PWR estimated from CCTF [Cylindrical Core Test Facility], SCTF [Slab Core Test Facility] and UPTF [Upper Plenum Test Facility] results

    International Nuclear Information System (INIS)

    Murao, Y.; Iguchi, T.; Sugimoto, J.

    1988-09-01

    The reflooding behavior in a PWR with a combined injection type ECCS was studied by comparing the test results from Cylindrical Core Test Facility (CCTF), Slab Core Test Facility (SCTF) and Upper Plenum Test Facility (UPTF). Core thermal-hydraulics is discussed mainly based on SCTF test data. In addition, the water accumulation behavior in hot legs and the break-through characteristics at tie plate are discussed

  7. Conceptual design study advanced concepts test (ACT) facility

    Energy Technology Data Exchange (ETDEWEB)

    Zaloudek, F.R.

    1978-09-01

    The Advanced Concepts Test (ACT) Project is part of program for developing improved power plant dry cooling systems in which ammonia is used as a heat transfer fluid between the power plant and the heat rejection tower. The test facility will be designed to condense 60,000 lb/hr of exhaust steam from the No. 1 turbine in the Kern Power Plant at Bakersfield, CA, transport the heat of condensation from the condenser to the cooling tower by an ammonia phase-change heat transport system, and dissipate this heat to the environs by a dry/wet deluge tower. The design and construction of the test facility will be the responsibility of the Electric Power Research Institute. The DOE, UCC/Linde, and the Pacific Northwest Laboratories will be involved in other phases of the project. The planned test facilities, its structures, mechanical and electrical equipment, control systems, codes and standards, decommissioning requirements, safety and environmental aspects, and energy impact are described. Six appendices of related information are included. (LCL)

  8. Safety requirements and safety experience of nuclear facilities in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Schnurer, H.L.

    1977-01-01

    Peaceful use of nuclear energy within the F.R.G. is rapidly growing. The Energy Programme of the Federal Government forecasts a capacity of up to 50.000 MW in 1985. Whereas most of this capacity will be of the LWR-Type, other activities are related to LMFBR - and HTGR - development, nuclear ships, and facilities of the nuclear fuel cycle. Safety of nuclear energy is the pacemaker for the realization of nuclear programmes and projects. Due to a very high population - and industrialisation density, safety has the priority before economical aspects. Safety requirements are therefore extremely stringent, which will be shown for the legal, the technical as well as for the organizational area. They apply for each nuclear facility, its site and the nuclear energy system as a whole. Regulatory procedures differ from many other countries, assigning executive power to state authorities, which are supervised by the Federal Government. Another particularity of the regulatory process is the large scope of involvement of independent experts within the licensing procedures. The developement of national safety requirements in different countries generates a necessity to collaborate and harmonize safety and radiation protection measures, at least for facilities in border areas, to adopt international standards and to assist nuclear developing countries. However, different nationally, regional or local situations might raise problems. Safety experience with nuclear facilities can be concluded from the positive construction and operation experience, including also a few accidents and incidents and the conclusions, which have been drawn for the respective factilities and others of similar design. Another tool for safety assessments will be risk analyses, which are under development by German experts. Final, a scope of future problems and developments shows, that safety of nuclear installations - which has reached a high performance - nevertheless imposes further tasks to be solved

  9. Heating facility for blanket and performance test

    Energy Technology Data Exchange (ETDEWEB)

    Furuya, Kazuyuki; Kuroda, Toshimasa; Enoeda, Mikio; Sato, Satoshi; Hatano, Toshihisa; Takatsu, Hideyuki; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Hara, Shigemitsu

    1999-03-01

    A design and a fabrication of heating test facility for a mock-up of the blanket module to be installed in International Thermonuclear Experimental Reactor (ITER) have been conducted to evaluate/demonstrate its heat removal performance and structural soundness under cyclic heat loads. To simulate surface heat flux to the blanket module, infrared heating method is adopted so as to heat large surface area uniformly. The infrared heater is used in vacuum environment (10{sup -4} Torr{approx}), and the lamps are cooled by air flowing through an annulus between the lamp and a cover tube made of quartz glass. Elastomer O rings (available to be used up to {approx}300degC) and used for vacuum seal at outer surface of the cover tube. To prevent excessive heating of the O ring, the end part of the cover tube is specially designed including the tube shape, flow path of air and gold coating on the surface of the cover tube to protect the O ring against thermal radiation from glowing tungsten filament. To examine the performance of the facility, steady state and cyclic operation of the infrared heater were conducted using a small-scaled shielding blanket mock-up as a test specimen. The important results are as follows: (1) Heat flux at the surface of the small-scaled mock-up measured by a calorimeter was {approx}0.2 MW/m{sup 2}. (2) A comparison of thermal analysis results and measured temperature responses showed that the small-scaled mock-up had good heat removal performance. (3) Steady state operation and cyclic operation with step response between the rated and zero powers of the infrared heater were successfully performed, and it was confirmed that this heating facility was well-prepared and available for the thermal cyclic test of a blanket module. (author)

  10. Large coil test facility conceptual design report

    International Nuclear Information System (INIS)

    Nelms, L.W.; Thompson, P.B.; Mann, T.L.

    1978-02-01

    In the development of a superconducting toroidal field (TF) magnet for The Next Step (TNS) tokamak reactor, several different TF coils, about half TNS size, will be built and tested to permit selection of a design and fabrication procedure for full-scale TNS coils. A conceptual design has been completed for a facility to test D-shaped TF coils, 2.5 x 3.5-m bore, operating at 4-6 K, cooled either by boiling helium or by forced-flow supercritical helium. Up to six coils can be accommodated in a toroidal array housed in a single vacuum tank. The principal components and systems in the facility are an 11-m vacuum tank, a test stand providing structural support and service connections for the coils, a liquid nitrogen system, a system providing helium both as saturated liquid and at supercritical pressure, coils to produce a pulsed vertical field at any selected test coil position, coil power supplies, process instrumentation and control, coil diagnostics, and a data acquisition and handling system. The test stand structure is composed of a central bucking post, a base structure, and two horizontal torque rings. The coils are bolted to the bucking post, which transmits all gravity loads to the base structure. The torque ring structure, consisting of beams between adjacent coils, acts with the bucking structure to react all the magnetic loads that occur when the coils are energized. Liquid helium is used to cool the test stand structure to 5 K to minimize heat conduction to the coils. Liquid nitrogen is used to precool gaseous helium during system cooldown and to provide thermal radiation shielding

  11. Model-based testing for software safety

    NARCIS (Netherlands)

    Gurbuz, Havva Gulay; Tekinerdogan, Bedir

    2017-01-01

    Testing safety-critical systems is crucial since a failure or malfunction may result in death or serious injuries to people, equipment, or environment. An important challenge in testing is the derivation of test cases that can identify the potential faults. Model-based testing adopts models of a

  12. Assessment of the facilities on Jackass Flats and other Nevada Test Site facilities for the new nuclear rocket program

    International Nuclear Information System (INIS)

    Chandler, G.; Collins, D.; Dye, K.; Eberhart, C.; Hynes, M.; Kovach, R.; Ortiz, R.; Perea, J.; Sherman, D.

    1992-01-01

    Recent NASA/DOE studies for the Space Exploration Initiative have demonstrated a critical need for the ground-based testing of nuclear rocket engines. Experience in the ROVER/NERVA Program, experience in the Nuclear Weapons Testing Program, and involvement in the new nuclear rocket program has motivated our detailed assessment of the facilities used for the ROVER/NERVA Program and other facilities located at the Nevada Test Site (NTS). The ROVER/NERVA facilities are located in the Nevada Research L, Development Area (NRDA) on Jackass Flats at NTS, approximately 85 miles northwest of Las Vegas. To guide our assessment of facilities for an engine testing program we have defined a program goal, scope, and process. To execute this program scope and process will require ten facilities. We considered the use of all relevant facilities at NTS including existing and new tunnels as well as the facilities at NRDA. Aside from the facilities located at remote sites and the inter-site transportation system, all of the required facilities are available at NRDA. In particular we have studied the refurbishment of E-MAD, ETS-1, R-MAD, and the interconnecting railroad. The total cost for such a refurbishment we estimate to be about $253M which includes additional contractor fees related to indirect, construction management, profit, contingency, and management reserves. This figure also includes the cost of the required NEPA, safety, and security documentation

  13. FMIT - the fusion materials irradiation test facility

    International Nuclear Information System (INIS)

    Liska, D.J.

    1980-01-01

    A joint effort by the Hanford Engineering Development Laboratory (HEDL) and Los Alamos Scientific Laboratory (LASL) has produced a preliminary design for a Fusion Materials Irradiation Test Facility (FMIT) that uses a high-power linear accelerator to fire a deuteron beam into a high-speed jet of molten lithium. The result is a continuous energy spectrum of neutrons with a 14-MeV average energy which can irradiate material samples to projected end-of-life levels in about 3 years, with a total accumulated fluence of 10 21 to 10 22 n/cm 2

  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. RADON-type disposal facility safety case for the co-ordinated research project on improvement of safety assessment methodologies for near surface radioactive waste disposal facilities (ISAM)

    International Nuclear Information System (INIS)

    Guskov, A.; Batanjieva, B.; Kozak, M.W.; Torres-Vidal, C.

    2002-01-01

    The ISAM safety assessment methodology was applied to RADON-type facilities. The assessments conducted through the ISAM project were among the first conducted for these kinds of facilities. These assessments are anticipated to lead to significantly improved levels of safety in countries with such facilities. Experience gained though this RADON-type Safety Case was already used in Russia while developing national regulatory documents. (author)

  16. Safety Research Experiment Facilities, Idaho National Engineering Laboratory, Idaho. Draft environmental statement

    International Nuclear Information System (INIS)

    1977-01-01

    This environmental statement was prepared in accordance with the National Environmental Policy Act of 1969 (NEPA) in support of the Energy Research and Development Administration's (ERDA) proposal for legislative authorization and appropriations 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 evalution of some design options and in the assessment of the long-term potential risk associated with wide-scale deployment of the FBR

  17. Lewis Research Center space station electric power system test facilities

    Science.gov (United States)

    Birchenough, Arthur G.; Martin, Donald F.

    1988-01-01

    NASA Lewis Research Center facilities were developed to support testing of the Space Station Electric Power System. The capabilities and plans for these facilities are described. The three facilities which are required in the Phase C/D testing, the Power Systems Facility, the Space Power Facility, and the EPS Simulation Lab, are described in detail. The responsibilities of NASA Lewis and outside groups in conducting tests are also discussed.

  18. Considerations on a PAHR test facility

    International Nuclear Information System (INIS)

    Boenisch, G.; Groetzbach, G.; Heinzel, V.; Kleefeld, K.; Kuechle, M.; Mueller, R.A.; Royl, P.; Schramm, K.; Smidt, D.; Werle, H.

    1976-01-01

    On the basis of a hypothetical core disruptive accident (HCDA) analysis the phenomena of the post accident phase are first identified which require experimental investigations and can only be studied in pile. Then the experimental requests for both debris bed and molten fuel pool studies are specified and grouped into three categories. For two of the categories the requests can be satisfied with loop experiments in thermal reactors. For the third category a 70 cm diameter test bed is needed and here the proposal is to use a flat core fast reactor with the test bed located below the core heated by axial leakage neutrons. Finally a conceptual design for such a reactor is presented where the test bed is loaded into an ex-vessel device and is removable on a carriage to a hot cell building. Maintenance and safety problems are briefly discussed and alternative solutions are mentioned

  19. Full scaled tests of the KERENA trademark containment cooling condenser at the INKA test facility

    International Nuclear Information System (INIS)

    Leyer, Stephan; Maisberger, Fabian; Lineva, Natalia; Wagner, Thomas; Doll, Mathias; Herbst, Vasilli; Wich, Michael

    2010-01-01

    KERENA trademark is a medium-capacity boiling water reactor. It combines passive safety systems with active safety equipment of service-proven design. The passive systems utilize basic laws of physics, such as gravity and natural convection, enabling them to function without any power supply or actuation by instrumentation and control (I and C) equipment. They are designed to bring the plant to a safe and stable condition without the aid of active systems. Furthermore, the passive safety features partially replace the active systems, which reduces costs significantly and provides a safe, reliable and economically competitive plant design. At the new test facility at Karlstein called INKA (Integral Test Stand Karlstein), the key components of the KERENA trademark passive safety concept - the Emergency Condenser (EC), the Containment Cooling Condenser (CCC) and the passive core flooding system (PCFS) - are presently under full-scale testing,. Integral system tests will also be performed to show how the passive safety systems interact under various anticipated accident conditions and to demonstrate the ability of the passive systems to bring the plant to a safe and stable condition without the aid of active systems or actuation by I and C signals. The passive pressure pulse transmitter (PPPT) will be included in these integral tests. In this report the experimental setup and the first test results with the full scaled Containment Cooling Condenser will be described. (orig.)

  20. Experience with the instrumentation tests in large sodium test facilities

    International Nuclear Information System (INIS)

    Lauhoff, Th.; Ruppert, E.; Stehle, H.; Vinzens, K.

    1976-01-01

    A facility is described for fast breeder core components (AKB) to test specially instrumented fuel dummies and blanket elements, and also absorber elements under simulated normal and extreme reactor conditions. In addition to endurance testing of a special sodium and high temperature sub-assembly, instrumentation is provided to investigate thermohydraulic and vibrational behaviour of core elements. During tests of > 3000 h at temperatures above 820 K the main sub-assembly characteristics, e.g. pressure drop, leakage flow, vibration and noise spectra can be reproduced. The use of eddy current flow meters, strain gauges, magnetostrictive noise sensors, pressure transducers, thermocouples, and acoustic surveillance devices, are described. (U.K.)

  1. Construction safety program for the National Ignition Facility Appendix A: Safety Requirements

    International Nuclear Information System (INIS)

    Cerruti, S.J.

    1997-01-01

    These rules apply to all LLNL employees, non-LLNL employees (including contract labor, supplemental labor, vendors, personnel matrixed/assigned from other National Laboratories, participating guests, visitors and students) and construction contractors/subcontractors. The General Safety and Health rules shall be used by management to promote accident prevention through indoctrination, safety and health training and on-the-job application. As a condition for contracts award, all contractors and subcontractors and their employees must certify on Form S ampersand H A-1 that they have read and understand, or have been briefed and understand, the National Ignition Facility OCIP Project General Safety Rules

  2. Construction safety program for the National Ignition Facility Appendix A: Safety Requirements

    Energy Technology Data Exchange (ETDEWEB)

    Cerruti, S.J.

    1997-01-14

    These rules apply to all LLNL employees, non-LLNL employees (including contract labor, supplemental labor, vendors, personnel matrixed/assigned from other National Laboratories, participating guests, visitors and students) and construction contractors/subcontractors. The General Safety and Health rules shall be used by management to promote accident prevention through indoctrination, safety and health training and on-the-job application. As a condition for contracts award, all contractors and subcontractors and their employees must certify on Form S & H A-1 that they have read and understand, or have been briefed and understand, the National Ignition Facility OCIP Project General Safety Rules.

  3. Consolidated Incineration Facility metals partitioning test

    International Nuclear Information System (INIS)

    Burns, D.B.

    1993-01-01

    Test burns were conducted at Energy and Environmental Research Corporation's rotary kiln simulator, the Solid Waste Incineration Test Facility, using surrogate CIF wastes spiked with hazardous metals and organics. The primary objective for this test program was measuring heavy metals partition between the kiln bottom ash, scrubber blowdown solution, and incinerator stack gas. Also, these secondary waste streams were characterized to determine waste treatment requirements prior to final disposal. These tests were designed to investigate the effect of several parameters on metals partitioning: incineration temperature; waste chloride concentration; waste form (solid or liquid); and chloride concentration in the scrubber water. Tests were conducted at three kiln operating temperatures. Three waste simulants were burned, two solid waste mixtures (paper, plastic, latex, and one with and one without PVC), and a liquid waste mixture (containing benzene and chlorobenzene). Toxic organic and metal compounds were spiked into the simulated wastes to evaluate their fate under various combustion conditions. Kiln offgases were sampled for volatile organic compounds (VOC), semi-volatile organic compounds (SVOC), polychlorinated dibenz[p]dioxins and polychlorinated dibenzofurans (PCDD/PCDF), metals, particulate loading and size distribution, HCl, and combustion products. Stack gas sampling was performed to determine additional treatment requirements prior to final waste disposal. Significant test results are summarized below

  4. Conceptual development of a test facility for spent fuel management

    Energy Technology Data Exchange (ETDEWEB)

    Park, S.W.; Lee, H.H.; Lee, J.Y.; Lee, J.S.; Ro, S.G. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    Spent fuel management is an important issue for nuclear power program, requiring careful planning and implementation. With the wait-and-see policy on spent fuel management in Korea, research efforts are directed at KAERI to develop advanced technologies for safer and more efficient management of the accumulating spent fuels. In support of these research perspectives, a test facility of pilot scale is being developed with provisions for integral demonstration of a multitude of technical functions required for spent fuel management. The facility, baptized SMART (Spent fuel MAnagement technology Research and Test facility), is to be capable of handling full size assembly of spent PWR fuel (as well as CANDU fuel) with a maximum capacity of 10 MTU/y (about 24 assemblies of PWR type). Major functions of the facility are consolidation of spent PWR fuel assembly into a half-volume package and optionally transformation of the fuel rod into a fuel of CANDU type (called DUPIC). Objectives of these functions are to demonstrate volume reduction of spent fuel (for either longer-term dry storage or direct disposal ) in the former case and direct refabrication of the spent PWR fuel into CANDU-type DUPIC fuel for reuse in CANDU reactors in the latter case, respectively. In addition to these major functions, there are other associated technologies to be demonstrated : such as waste treatment, remote maintenance, safeguards, etc. As the facility is to demonstrate not only the functional processes but also the safety and efficiency of the test operations, engineering criteria equivalent to industrial standards are incorporated in the design concept. The hot cell structure enclosing the radioactive materials is configured in such way to maximize costs within the given functional and operational requirements. (author). 3 tabs., 4 figs.

  5. Conceptual development of a test facility for spent fuel management

    International Nuclear Information System (INIS)

    Park, S.W.; Lee, H.H.; Lee, J.Y.; Lee, J.S.; Ro, S.G.

    1997-01-01

    Spent fuel management is an important issue for nuclear power program, requiring careful planning and implementation. With the wait-and-see policy on spent fuel management in Korea, research efforts are directed at KAERI to develop advanced technologies for safer and more efficient management of the accumulating spent fuels. In support of these research perspectives, a test facility of pilot scale is being developed with provisions for integral demonstration of a multitude of technical functions required for spent fuel management. The facility, baptized SMART (Spent fuel MAnagement technology Research and Test facility), is to be capable of handling full size assembly of spent PWR fuel (as well as CANDU fuel) with a maximum capacity of 10 MTU/y (about 24 assemblies of PWR type). Major functions of the facility are consolidation of spent PWR fuel assembly into a half-volume package and optionally transformation of the fuel rod into a fuel of CANDU type (called DUPIC). Objectives of these functions are to demonstrate volume reduction of spent fuel (for either longer-term dry storage or direct disposal ) in the former case and direct refabrication of the spent PWR fuel into CANDU-type DUPIC fuel for reuse in CANDU reactors in the latter case, respectively. In addition to these major functions, there are other associated technologies to be demonstrated : such as waste treatment, remote maintenance, safeguards, etc. As the facility is to demonstrate not only the functional processes but also the safety and efficiency of the test operations, engineering criteria equivalent to industrial standards are incorporated in the design concept. The hot cell structure enclosing the radioactive materials is configured in such way to maximize costs within the given functional and operational requirements. (author). 3 tabs., 4 figs

  6. Safety assessment of near surface radioactive waste disposal facilities: Model intercomparison using simple hypothetical data (Test Case 1). First report of NSARS. Part of the co-ordinated research programme on the safety assessment of near surface radioactive waste disposal facilities (NSARS)

    International Nuclear Information System (INIS)

    1995-11-01

    In many countries near surface disposal is the preferred option for the comparatively large volumes of low and intermediate level wastes which arise during nuclear power plant operations, nuclear fuel reprocessing and also for the wastes arising from radionuclide applications in hospitals and research establishments. Near surface disposal is also widely practised in the case of hazardous wastes from chemical industries. It is obviously necessary to show that waste disposal methods are safe and that both man and the environment will be adequately protected. Following a previous related Co-ordinated Research Programme (CRP) on ''Migration and Biological Transfer of Radionuclides from Shallow Land Burial'' during 1985 to 1989 (IAEA-TECDOC-579, Vienna, 1990), the issue of reliability of safety assessments was identified as an important topic for further support and development. A new CRP was formulated with the acronym NSARS (Near Surface Radioactive Waste Disposal Safety Assessment Reliability Study). This technical document is the first report of from the CRP and contains the intercomparison of results of the first test exercise (Test Case 1) on modelling of potential radiation exposures as a result of near surface disposal. Test Case 1 is based on entirely hypothetical data and includes consideration of exposures due to leaching and as a result of human intrusion into the site. Refs, figs and tabs

  7. Safety assessment of near surface radioactive waste disposal facilities: Model intercomparison using simple hypothetical data (Test Case 1). First report of NSARS. Part of the co-ordinated research programme on the safety assessment of near surface radioactive waste disposal facilities (NSARS)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-11-01

    In many countries near surface disposal is the preferred option for the comparatively large volumes of low and intermediate level wastes which arise during nuclear power plant operations, nuclear fuel reprocessing and also for the wastes arising from radionuclide applications in hospitals and research establishments. Near surface disposal is also widely practised in the case of hazardous wastes from chemical industries. It is obviously necessary to show that waste disposal methods are safe and that both man and the environment will be adequately protected. Following a previous related Co-ordinated Research Programme (CRP) on ``Migration and Biological Transfer of Radionuclides from Shallow Land Burial`` during 1985 to 1989 (IAEA-TECDOC-579, Vienna, 1990), the issue of reliability of safety assessments was identified as an important topic for further support and development. A new CRP was formulated with the acronym NSARS (Near Surface Radioactive Waste Disposal Safety Assessment Reliability Study). This technical document is the first report of from the CRP and contains the intercomparison of results of the first test exercise (Test Case 1) on modelling of potential radiation exposures as a result of near surface disposal. Test Case 1 is based on entirely hypothetical data and includes consideration of exposures due to leaching and as a result of human intrusion into the site. Refs, figs and tabs.

  8. Mirror fusion test facility plasma diagnostics system

    International Nuclear Information System (INIS)

    Thomas, S.R. Jr.; Coffield, F.E.; Davis, G.E.; Felker, B.

    1979-01-01

    During the past 25 years, experiments with several magnetic mirror machines were performed as part of the Magnetic Fusion Energy (MFE) Program at LLL. The latest MFE experiment, the Mirror Fusion Test Facility (MFTF), builds on the advances of earlier machines in initiating, stabilizing, heating, and sustaining plasmas formed with deuterium. The goals of this machine are to increase ion and electron temperatures and show a corresponding increase in containment time, to test theoretical scaling laws of plasma instabilities with increased physical dimensions, and to sustain high-beta plasmas for times that are long compared to the energy containment time. This paper describes the diagnostic system being developed to characterize these plasma parameters

  9. Considerations in the safety assessment of sealed nuclear facilities

    International Nuclear Information System (INIS)

    1991-06-01

    This report is a part of the International Atomic Energy Agency's radioactive waste management programme, whose objective is to provide assistance to Member States in developing guidance for identifying safe alternatives for isolating radioactive waste from man and his environment. This report attempts to integrate information from the previous reports on decommissioning of nuclear facilities, mitigation of accidents at such facilities, and performance assessment of disposal systems to provide useful advice and qualitative guidance to those responsible for performance and safety assessments of sealed nuclear facilities by giving an overview of possible approaches and techniques for such assessments. In this context, the establishment of requirements and rules governing the radiological safety of personnel, the general public, and the environment for sealing and post-sealing activities will enable the choice of the most appropriated approach and help to promote consistency in both decommissioning and waste management standards. The near-field effects discussed in this document include gas generation, interactions of the groundwater and the residual water with other components of the system, thermal, thermo-mechanical, radiation effects and chemical and geochemical reactions. 59 refs, figs and tabs

  10. Development of new irradiation facility for BWR safety research

    International Nuclear Information System (INIS)

    Okada, Yuji; Magome, Hirokatsu; Iida, Kazuhiro; Hanawa, Hiroshi; Ohmi, Masao

    2013-01-01

    In JAEA (Japan Atomic Energy Agency), about the irradiation embrittlement of the reactor pressure vessel and the stress corrosion cracking of reactor core composition apparatus concerning the long-term use of the light water reactor (BWR), in order to check the influence of the temperature, pressure, and water quality, etc on BWR condition. The water environmental control facility which performs irradiation assisted stress corrosion-cracking (IASCC) evaluation under BWR irradiation environment was fabricated in JMTR (Japan Materials Testing Reactor). This report is described the outline of manufacture of the water environmental control facility for doing an irradiation test using the saturation temperature capsule after JMTR re-operation. (author)

  11. Safety report for Central Interim Storage facility for radioactive waste from small producers

    International Nuclear Information System (INIS)

    Zeleznik, N.; Mele, I.

    2004-01-01

    protection and remediation of minor defects in the facility. The safety report also incorporates the implementation projects for the inventory rearrangement and the projected waste quantities up to 2013, when the LILW repository should be available. The most important part of the safety report deals with safety analyses of all relevant operational conditions which could influence safety of the storage and its inventory, taking into account also the phase of reconstruction works. The Safety Report is prepared in accordance with the requirements of Regulation E2 (Pravilnik E2) and The Act on Ionising Radiation Protection and Nuclear Safety from 2002. After defining the objective of the safety report, the chapters deal with the following topics: The safety approach to the LILW storage, Description and location analysis of the Central interim storage, Technical characteristic of the Central interim storage, Safety analysis of the Central interim storage, Organizational measures for construction and pre-operational testing, Organizational measures for commissioning and normal operation of the Central interim storage, Operational conditions and limitations, Ionizing radiation protection service, its methods and equipment, Radioactive waste management and disposal, Review of the plans, measures and procedures to prevent radiological accidents, Quality assurance programme, Review of the measures for physical protection of the LILW storage and stored radioactive waste, Planned measures and necessary equipment for closure of the Central interim storage. The details on preparation of the Safety report, including the dilemmas and comments on different chapters, are presented in this paper. The most important findings of the safety analysis are also given.(author)

  12. Final safety analysis report (FSAR) for waste receiving and processing (WRAP) facility

    International Nuclear Information System (INIS)

    Weidert, J.R.

    1997-01-01

    This safety analysis report provides a summary description of the WRAP Facility, focusing on significant safety-related characteristics of the location and facility design. This report demonstrates that adherence to the safety basis wi11 ensure necessary operational safety considerations have been addressed sufficiently and justifies the adequacy of the safety basis in protecting the health and safety of the public, workers, and the environment

  13. Organization and staffing of the regulatory body for nuclear facilities. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    The purpose of this safety guide is to provide recommendations for national authorities on the appropriate management system, organization and staffing for the regulatory body responsible for the regulation of nuclear facilities in order to achieve compliance with the applicable safety requirements. This safety guide covers the organization and staffing in relation to nuclear facilities such as: enrichment and fuel manufacturing plants. Nuclear power plants. Other reactors such as research reactors and critical assemblies. Spent fuel reprocessing plants. And radioactive waste management facilities such as treatment, storage and disposal facilities. This safety guide also covers issues related to the decommissioning of nuclear facilities, the closure of waste disposal facilities and site rehabilitation

  14. Sodium-water reaction test facility (SWAT-3)

    International Nuclear Information System (INIS)

    Shimazu, Hisashi; Ukechi, Kazutoshi; Sasakura, Kazutake; Kusunoki, Junichi

    1976-01-01

    In the development of the liquid metal cooled fast breeder reactor (LMFBR), the steam generator (SG) is considered one of the most important components. The Power Reactor and Nuclear Fuel Development Corporation (PNC) is now promoting the research and development of the SG system used with the prototype fast breeder reactor ''Monju''. In this research, the phenomena of the sodium-water reaction in the SG are the key which must be investigated for the solution of problems. The test facility (SWAT-3) simulating Monju's SG on the scale of 1/2.5 was designed, fabricated and installed by IHI at Oarai Engineering Center of PNC, its pre-operation being accomplished in February 1975. The purpose of SWAT-3 is summarized as follows: (1) To perform an overall test on the safety of Monju's SG and intermediate heat transport system under the design condition against sodium-water reaction accidents. (2) To investigate the damage of the SG structure caused by the sodium-water reaction, and the possibility of repair and recovery operations. The first test was accomplished successfully on June 9, 1975. As a result of the test, the fundamental function of this test facility was proven to be satisfactory as expected. (auth.)

  15. Fuels and materials testing capabilities in Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Baker, R.B.; Chastain, S.A.; Culley, G.E.; Ethridge, J.L.; Lovell, A.J.; Newland, D.J.; Pember, L.A.; Puigh, R.J.; Waltar, A.E.

    1989-01-01

    The Fast Flux Test Facility (FFTF) reactor, which started operating in 1982, is a 400 MWt sodium-cooled fast neutron reactor located in Hanford, Washington State, and operated by Westinghouse Hanford Co. under contract with U.S. Department of Energy. The reactor has a wide variety of functions for irradiation tests and special tests, and its major purpose is the irradiation of fuel and material for liquid metal reactor, nuclear reactor and space reactor projects. The review first describes major technical specifications and current conditions of the FFTF reactor. Then the plan for irradiation testing is outlined focusing on general features, fuel pin/assembly irradiation tests, and absorber irradiation tests. Assemblies for special tests include the material open test assembly (MOTA), fuel open test assembly (FOTA), closed loop in-reactor assembly (CLIRA), and other special fuel assemblies. An interim examination and maintenance cell (FFTF/IEM cell) and other hot cells are used for nondestructive/destructive tests and physical/mechanical properties test of material after irradiation. (N.K.)

  16. The technological safety in facilities that manage radioactive sources

    International Nuclear Information System (INIS)

    Lizcano, D.

    2014-10-01

    The sealed radioactive sources are used inside a wide range of applications in the medicine, industry and investigation around the world. These sources can contain a great radionuclides variety, exhibiting a wide spectrum of activities and radiological half lives. This way, we can find pattern sources of radionuclides as Americium-241, Plutonium-238, Plutonium-239, Thorium-228 and Thorium-230, etc., with some activity of kBq in research laboratories, Iridium-192 and Cesium-137 sources used in brachytherapy with GBq activities, until sources with P Bq activities in industrial irradiators of Cobalt-60 and Cesium-137. This document approach the physical safety that entities like the IAEA recommends for the facilities that contain sealed sources, especially the measures that are taking in the Instituto Nacional de Investigaciones Nucleares (ININ) and others government facilities. (Author)

  17. 40 CFR 160.31 - Testing facility management.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 23 2010-07-01 2010-07-01 false Testing facility management. 160.31... GOOD LABORATORY PRACTICE STANDARDS Organization and Personnel § 160.31 Testing facility management. For each study, testing facility management shall: (a) Designate a study director as described in § 160.33...

  18. 40 CFR 792.31 - Testing facility management.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 31 2010-07-01 2010-07-01 true Testing facility management. 792.31... facility management. For each study, testing facility management shall: (a) Designate a study director as... appropriately tested for identity, strength, purity, stability, and uniformity, as applicable. (e) Assure that...

  19. Module Testing Techniques for Nuclear Safety Critical Software Using LDRA Testing Tool

    International Nuclear Information System (INIS)

    Moon, Kwon-Ki; Kim, Do-Yeon; Chang, Hoon-Seon; Chang, Young-Woo; Yun, Jae-Hee; Park, Jee-Duck; Kim, Jae-Hack

    2006-01-01

    The safety critical software in the I and C systems of nuclear power plants requires high functional integrity and reliability. To achieve those requirement goals, the safety critical software should be verified and tested according to related codes and standards through verification and validation (V and V) activities. The safety critical software testing is performed at various stages during the development of the software, and is generally classified as three major activities: module testing, system integration testing, and system validation testing. Module testing involves the evaluation of module level functions of hardware and software. System integration testing investigates the characteristics of a collection of modules and aims at establishing their correct interactions. System validation testing demonstrates that the complete system satisfies its functional requirements. In order to generate reliable software and reduce high maintenance cost, it is important that software testing is carried out at module level. Module testing for the nuclear safety critical software has rarely been performed by formal and proven testing tools because of its various constraints. LDRA testing tool is a widely used and proven tool set that provides powerful source code testing and analysis facilities for the V and V of general purpose software and safety critical software. Use of the tool set is indispensable where software is required to be reliable and as error-free as possible, and its use brings in substantial time and cost savings, and efficiency

  20. [RADIATION SAFETY DURING REMEDIATION OF THE "SEVRAO" FACILITIES].

    Science.gov (United States)

    Shandala, N K; Kiselev, S M; Titov, A V; Simakov, A V; Seregin, V A; Kryuchkov, V P; Bogdanova, L S; Grachev, M I

    2015-01-01

    Within a framework of national program on elimination of nuclear legacy, State Corporation "Rosatom" is working on rehabilitation at the temporary waste storage facility at Andreeva Bay (Northwest Center for radioactive waste "SEVRAO"--the branch of "RosRAO"), located in the North-West of Russia. In the article there is presented an analysis of the current state of supervision for radiation safety of personnel and population in the context of readiness of the regulator to the implementation of an effective oversight of radiation safety in the process of radiation-hazardous work. Presented in the article results of radiation-hygienic monitoring are an informative indicator of the effectiveness of realized rehabilitation measures and characterize the radiation environment in the surveillance zone as a normal, without the tendency to its deterioration.

  1. Radiologic safety program for ionizing radiation facilities in Parana, Brazil

    International Nuclear Information System (INIS)

    Schmidt, M.F.S.; Tilly Junior, J.G.

    1997-01-01

    A radiologic safety program for inspection, licensing and control of the use of ionizing radiation in medical, industrial and research facilities in Parana, Brazil is presented. The program includes stages such as: 1- division into implementation phases considering the activity development for each area; 2-use of the existing structure to implement and to improve services. The development of the program will permit to evaluate the improvement reached and to correct operational strategic. As a result, a quality enhancement at the services performed, a reduction for radiation dose exposure and a faster response for emergency situations will be expected

  2. Cylindrical core reflood test facility (CCTF) and slab core reflood test facility (SCTF) for Japan Atomic Energy Research Institute (JAERI)

    International Nuclear Information System (INIS)

    1981-01-01

    IHI has designed and constructed the CCTF at JAERI to be used in the safety analysis research on the loss of coolant accident in a PWR plant. This test facility is planned so that reflood phenomenon in the PWR plant (a phenomenon is that the bared and overheated core is reflooded by the emergency core cooling system when the coolant loss accident occurred) is simulated under various test conditions. The CCTF is the largest-scale test plant in the world, composed of approximately 2000 simulated fuel rods (electric heaters), 1 simulated pressure vessel, 4 primary cooling loops, 2 simulated steam generators, emergency core cooling system, and so on. The test conditions are controlled, and the test steps are sequentially progressed by the computing system, and test data are collected by the data acquisition system. Furthermore, IHI is now designing and constructing the SCTF in accordance with the JAERI research plan. The SCTF is similar to the CCTF in scale. Main feature of the SCTF is the form of the simulated core and the simulated pressure vessel, which is of slab construction to be representative of the radial section of the PWR reactor. Reliable and various data for safety analysis are expected by the CCTF and the SCTF. (author)

  3. Nuclear Safety Co-Ordination within Oak Ridge Operations Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, W. A.; Pryor, W. A. [Research and Development Division, United States Atomic Energy Commission, Oak Ridge, TN (United States)

    1966-05-15

    The Oak Ridge Operations Office of the USAEC has within its jurisdiction multiple contractors and facilities for research and for the production of fissile materials for the atomic energy programme. Among these facilities are gaseous diffusion plants for the production of {sup 235}U-enriched uranium hexafluoride, plants for the fabrication of special components and fuel for research and production reactors, and laboratories for pilot plant studies and basic research in nuclear technology. One research laboratory is also actively engaged in criticality experimental programmes and has been a major contributor of criticality data for safety applications. These diversified programmes include the processing, fabrication and transport of practically all forms and isotopic enrichments of uranium in quantities commensurate with both laboratory and volume production requirements. Consequently, adequate nuclear safety control with reasonable economy for operations of this magnitude demands not only co-ordination and liaison between contractor and USAEC staffs, but a continuing reappraisal of safety applications in light of the most advanced information. This report outlines the role of the Oak Ridge Operations Office in these pursuits and describes as examples some specific problems in which this office co-ordinated actions necessary for their resolution. Other examples are given of parametric and procedural applications in plant processes and fissile shipments emphasizing the use of recent experimental or calculated data. These examples involve the use of mass and geometric variables, neutron absorbers and moderation control. Departures from limits specified in existing nuclear safety guides are made to advantage in light of new data, special equipment design, contingencies and acceptable risks. (author)

  4. 242-A Evaporator crystallizer facility integrated annual safety appraisal

    International Nuclear Information System (INIS)

    1991-01-01

    This report provides the results of the Fiscal Year (FY) 1991 Annual Integrated Safety Appraisal of the 242-A Evaporator Crystallizer Facility in the Hanford 200 East Area. The appraisal was conducted in December 1990 and January 1991, by the Waste Tank Safety Assurance (WTSA) organizations in conjunction with Radiological Engineering, Criticality Safety, Packaging and Shipping Safety, Emergency Preparedness, Environmental Compliance, and Quality Assurance. Reports of these eight organizations are presented as Sections 2 through 7 of this report. The purpose of the appraisal was to verify that the 242-A Evaporator meets US Department of Energy (DOE) and Westinghouse Hanford Company (WHC) requirements and current industry standards of good practice for the areas being appraised. A further purpose was to identify areas in which program effectiveness could be improved. In accordance with the guidance of WHC Management Requirements and Procedures (MRP)5.6, previously identified deficiencies which are being resolved by line management were not repeated as Findings or Observations unless progress or intended disposition was considered to be unsatisfactory

  5. 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

  6. Fast Flux Test Facility core system

    International Nuclear Information System (INIS)

    Ethridge, J.L.; Baker, R.B.; Leggett, R.D.; Pitner, A.L.; Waltar, A.E.

    1990-11-01

    A review of Liquid Metal Reactor (LMR) core system accomplishments provides an excellent road map through the maze of issues that faced reactor designers 10 years ago. At that time relatively large uncertainties were associated with fuel pin and fuel assembly performance, irradiation of structural materials, and performance of absorber assemblies. The extensive core systems irradiation program at the US Department of Energy's Fast Flux Test Facility (FFTF) has addressed each of these principal issues. As a result of the progress made, the attention of long-range LMR planners and designers can shift away from improving core systems and focus on reducing capital costs to ensure the LMR can compete economically in the 21st century with other nuclear reactor concepts. 3 refs., 6 figs., 1 tab

  7. The CERN linear collider test facility (CTF)

    International Nuclear Information System (INIS)

    Baconnier, Y.; Battisti, S.; Bossart, R.; Delahaye, J.P.; Geissler, K.K.; Godot, J.C.; Huebner, K.; Madsen, J.H.B.; Potier, J.P.; Riche, A.J.; Sladen, J.; Suberlucq, G.; Wilson, I.; Wuensch, W.

    1992-01-01

    The CTF (Collider Test Facility) was brought into service last year. The 3 GHz gun produced a beam of 3 MeV/c which was accelerated to 40 MeV/c. This beam, passing a prototype CLIC (linear collider) structure, generated a sizeable amount of 30 GHz power. This paper describes the results and experience with the gun driven by a 8 ns long laser pulse and its CsI photo cathode, the beam behaviour, the beam diagnostics in particular with the bunch measurements by Cerenkov or transition radiation light and streak camera, the photo cathode research, and the beam dynamics studies on space charge effects. (Author)4 figs., tab., 6 refs

  8. Counterpart experimental study of ISP-42 PANDA tests on PUMA facility

    International Nuclear Information System (INIS)

    Yang, Jun; Choi, Sung-Won; Lim, Jaehyok; Lee, Doo-Yong; Rassame, Somboon; Hibiki, Takashi; Ishii, Mamoru

    2013-01-01

    Highlights: ► Counterpart tests were performed on two large-scale BWR integral facilities. ► Similarity of post-LOCA system behaviors observed between two tests. ► Passive core and containment cooling systems work as design in both tests. -- Abstract: A counterpart test to the Passive Nachwärmeabfuhr und Druckabbau Test Anlage (Passive Decay Heat Removal and Depressurization Test Facility, PANDA) International Standard Problem (ISP)-42 test was conducted at the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) facility. Aimed to support code validation on a range of light water reactor (LWR) containment issues, the ISP-42 test consists of six sequential phases (Phases A–F) with separately defined initial and boundary conditions, addressing different stages of anticipated accident scenario and system responses. The counterpart test was performed from Phases A to D, which are within the scope of the normal integral tests performed on the PUMA facility. A scaling methodology was developed by using the PANDA facility as prototype and PUMA facility as test model, and an engineering scaling has been applied to the PUMA facility. The counterpart test results indicated that functions of passive safety systems, such as passive containment cooling system (PCCS) start-up, gravity-driven cooling system (GDCS) discharge, PCCS normal operation and overload function were confirmed in both the PANDA and PUMA facilities with qualitative similarities

  9. Design, Evaluation and Test Technology Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The mission of this facility, which is composed of numerous specialized facilities, is to provide capabilities to simulate a wide range of environments for component...

  10. 10-MWe pilot-plant-receiver panel test requirements document solar thermal test facility

    Energy Technology Data Exchange (ETDEWEB)

    1978-08-25

    Testing plans for a full-scale test receiver panel and supporting hardware which essentially duplicate both physically and functionally, the design planned for the Barstow Solar Pilot Plant are presented. Testing is to include operation during normal start and shutdown, intermittent cloud conditions, and emergencies to determine the panel's transient and steady state operating characteristics and performance under conditions equal to or exceeding those expected in the pilot plant. The effects of variations of input and output conditions on receiver operation are also to be investigated. Test hardware are described, including the pilot plant receiver, the test receiver assembly, receiver panel, flow control, electrical control and instrumentation, and structural assembly. Requirements for the Solar Thermal Test Facility for the tests are given. The safety of the system is briefly discussed, and procedures are described for assembly, installation, checkout, normal and abnormal operations, maintenance, removal and disposition. Also briefly discussed are quality assurance, contract responsibilities, and test documentation. (LEW)

  11. Plan for 3-D full-scale earthquake testing facility

    International Nuclear Information System (INIS)

    Ohtani, K.

    2001-01-01

    Based on the lessons learnt from the Great Hanshin-Awaji Earthquake, National Research Institute for Earth Science and Disaster Prevention plan to construct the 3-D Full-Scale Earthquake Testing Facility. This will be the world's largest and strongest shaking table facility. This paper describes the outline of the project for this facility. This facility will be completed in early 2005. (author)

  12. Information on the Advanced Plant Experiment (APEX) Test Facility

    International Nuclear Information System (INIS)

    Smith, Curtis Lee

    2015-01-01

    The purpose of this report provides information related to the design of the Oregon State University Advanced Plant Experiment (APEX) test facility. Information provided in this report have been pulled from the following information sources: Reference 1: R. Nourgaliev and et.al, 'Summary Report on NGSAC (Next-Generation Safety Analysis Code) Development and Testing,' Idaho National Laboratory, 2011. Note that this is report has not been released as an external report. Reference 2: O. Stevens, Characterization of the Advanced Plant Experiment (APEX) Passive Residual Heat Removal System Heat Exchanger, Master Thesis, June 1996. Reference 3: J. Reyes, Jr., Q. Wu, and J. King, Jr., Scaling Assessment for the Design of the OSU APEX-1000 Test Facility, OSU-APEX-03001 (Rev. 0), May 2003. Reference 4: J. Reyes et al, Final Report of the NRC AP600 Research Conducted at Oregon State University, NUREG/CR-6641, July 1999. Reference 5: K. Welter et al, APEX-1000 Confirmatory Testing to Support AP1000 Design Certification (non-proprietary), NUREG-1826, August 2005.

  13. W-1 Sodium Loop Safety Facility experiment centerline fuel thermocouple performance

    International Nuclear Information System (INIS)

    Meyers, S.C.; Henderson, J.M.

    1980-05-01

    The W-1 Sodium Loop Safety Facility (SLSF) experiment is the fifth in a series of experiments sponsored by the Department of Energy (DOE) as part of the National Fast Breeder Reactor (FBR) Safety Assurance Program. The experiments are being conducted under the direction of Argonne National Laboratory (ANL) and Hanford Engineering Development Laboratory (HEDL). The irradiation phase of the W-1 SLSF experiment was conducted between May 27 and July 20, 1979, and terminated with incipient fuel pin cladding failure during the final boiling transient. Experimental hardware and facility performed as designed, allowing completion of all planned tests and test objectives. This paper focuses on high temperature in-fuel thermocouples and discusses their development, fabrication, and performance in the W-1 experiment

  14. Construction safety program for the National Ignition Facility

    International Nuclear Information System (INIS)

    Cerruti, S.J.

    1997-01-01

    The Construction Safety Program (CSP) for NIF sets forth the responsibilities, guidelines, rules, policies and regulations for all workers involved in the construction, special equipment installation, acceptance testing, and initial activation and operation of NIF at LLNL during the construction period of NIF

  15. Construction safety program for the National Ignition Facility

    Energy Technology Data Exchange (ETDEWEB)

    Cerruti, S.J.

    1997-06-26

    The Construction Safety Program (CSP) for NIF sets forth the responsibilities, guidelines, rules, policies and regulations for all workers involved in the construction, special equipment installation, acceptance testing, and initial activation and operation of NIF at LLNL during the construction period of NIF.

  16. Preservation of FFTF Data Related to Passive Safety Testing

    International Nuclear Information System (INIS)

    Wootan, David W.; Butner, R. Scott; Omberg, Ronald P.; Makenas, Bruce J.; Nielsen, Deborah L.

    2010-01-01

    One of the goals of the Fuel Cycle Research and Development Program (FCRD) is to preserve the knowledge that has been gained in the United States on Liquid Metal Reactors (LMR). A key area deserving special attention for preservation is the data relating to passive safety testing that was conducted in FFTF and EBR-II during the 1980's. Accidents at Unit 4 of the Chernobyl Station and Unit 2 at Three Mile Island changed the safety paradigm of the nuclear power industry. New emphasis was placed on assured safety based on intrinsic plant characteristics that protect not only the public, but the significant investment in the plant as well. Plants designated to perform in this manner are considered to be passively safe since no active sensor/alarm system or human intervention is required to bring the reactor to a safe shutdown condition. The liquid metal reactor (LMR) has several key characteristics needed for a passively safe reactor: reactor coolant with superior heat transfer capability and very high boiling point, low (atmospheric) system pressures, and reliable negative reactivity feedback. The credibility of the design for a passively safe LMR rests on two issues: the validity of analytic methods used to predict passive safety performance and the availability of relevant test data to calibrate design tools. Safety analysis methods used to analyze LMRs under the old safety paradigm were focused on calculating the source term for the Core Disruptive Accident. Passive safety design requires refined analysis methods for transient events because treatment of the detailed reactivity feedbacks is important in predicting the response of the reactor. Similarly, analytic tools should be calibrated against actual test experience in existing LMR facilities. The principal objectives of the combined FFTF natural circulation and Passive Safety Testing program were: (1) to verify natural circulation as a reliable means to safely remove decay heat, (2) to extend passive safety

  17. Construction safety program for the National Ignition Facility

    International Nuclear Information System (INIS)

    Cerruti, S.J.

    1997-01-01

    The Construction Safety Program (CSP) for NIF sets forth the responsibilities, guidelines, rules, policies and regulations for all workers involved in the construction, special equipment installation, acceptance testing, and initial activation and operation of NIF at LLNL during the construction period of NIF. During this period, all workers are required to implement measures to create a universal awareness which promotes safe practice at the work site, and which will achieve NIF's management objectives in preventing accidents and illnesses. Construction safety for NIF is predicated on everyone performing their jobs in a manner which prevents job-related disabling injuries and illnesses. The CSP outlines the minimum environment, safety, and health (ES ampersand H) standards, LLNL policies and the Construction Industry Institute (CII) Zero Injury Techniques requirements that all workers at the NIF construction site shall adhere to during the construction period of NIF. It identifies the safety requirements which the NIF organizational Elements, construction contractors and construction subcontractors must include in their safety plans for the construction period of NIF, and presents safety protocols and guidelines which workers shall follow to assure a safe and healthful work environment. The CSP also identifies the ES ampersand H responsibilities of LLNL employees, non-LLNL employees, construction contractors, construction subcontractors, and various levels of management within the NIF Program at LLNL. In addition, the CSP contains the responsibilities and functions of ES ampersand H support organizations and administrative groups, and describes their interactions with the NIF Program

  18. Construction safety program for the National Ignition Facility

    Energy Technology Data Exchange (ETDEWEB)

    Cerruti, S.J.

    1997-01-01

    The Construction Safety Program (CSP) for NIF sets forth the responsibilities, guidelines, rules, policies and regulations for all workers involved in the construction, special equipment installation, acceptance testing, and initial activation and operation of NIF at LLNL during the construction period of NIF. During this period, all workers are required to implement measures to create a universal awareness which promotes safe practice at the work site, and which will achieve NIF`s management objectives in preventing accidents and illnesses. Construction safety for NIF is predicated on everyone performing their jobs in a manner which prevents job-related disabling injuries and illnesses. The CSP outlines the minimum environment, safety, and health (ES&H) standards, LLNL policies and the Construction Industry Institute (CII) Zero Injury Techniques requirements that all workers at the NIF construction site shall adhere to during the construction period of NIF. It identifies the safety requirements which the NIF organizational Elements, construction contractors and construction subcontractors must include in their safety plans for the construction period of NIF, and presents safety protocols and guidelines which workers shall follow to assure a safe and healthful work environment. The CSP also identifies the ES&H responsibilities of LLNL employees, non-LLNL employees, construction contractors, construction subcontractors, and various levels of management within the NIF Program at LLNL. In addition, the CSP contains the responsibilities and functions of ES&H support organizations and administrative groups, and describes their interactions with the NIF Program.

  19. 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.

  20. New safety performance indicators for safety assessment of radioactive waste disposal facilities. Cuban experience

    International Nuclear Information System (INIS)

    Peralta Vital, J.L.; Castillo, R.G.; Olivera, J.

    2002-01-01

    The paper shows the Cuban experience on implementing geological disposal of radioactive waste and the necessity for identifying new safety performance indicators for the safety assessment (SA) of radioactive waste disposal facilities. The selected indicator was the concentration of natural radioactive elements (U, Ra, Th, K) in the Cuban geologic environment. We have carried out a group of investigations, which have allowed characterising the concentration for the whole Country, creating a wide database where this indicator is associated with the lithology. The main lithologies in Cuba are: the sedimentary rocks (70 percent of national occurrence), which are present in the three regions (limestone and lutite), and finally the igneous and metamorphic rocks. The results show the concentrations ranges of the natural radionuclides associated fundamentally to the variation in the lithology and geographical area of the Country. In Cuba, the higher concentration (ppm) of Uranium and Radium are referenced to the Central region associated to Skarn, while for Thorium (ppm) and Potassium (%), in the East region the concentration peaks in Tuffs have been found. The concentrations ranges obtained are preliminary, they characterise the behaviour of this parameter for the Cuban geology, but they do not represent limits for safety assessment purposes yet. Also other factors should be taken into account as the assessment context, time scales and others assumptions before establishing the final concentration limits for the natural radionuclides as a radiological and nuclear safety performance indicator complementary to dose and risk for safety assessment for radiological and nuclear facilities. (author)