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

Validation experiments of nuclear characteristics of the fast-thermal system HERBE

International Nuclear Information System (INIS)

Pesic, M.; Zavaljevski, N.; Marinkovic, P.; Stefanovis, D.; Nikolic, D.; Avdic, S.

1992-01-01

In 1988/90 a coupled fast-thermal system HERBE at RB reactor, based on similar facilities, is designed and realized. Fast core of HERBE is built of natural U fuel in RB reactor center surrounded by the neutron filter and neutron converter located in an independent Al tank. Fast zone is surrounded by thermal neutron core driver. Designed nuclear characteristics of HERBE core are validated in the experiments described in the paper. HERBE cell parameters were calculated with developed computer codes: VESNA and DENEB. HERBE system criticality calculation are performed with 4G 2D RZ computer codes GALER and TWENTY GRAND, 1D multi-group AVERY code and 3D XYZ few-group TRITON computer code. The experiments for determination of critical level, dρ/dH, and reactivity of safety rods are accomplished in order to validate calculation results. Specific safety experiment is performed in aim to determine reactivity of flooded fast zone in possible accident. A very good agreements with calculation results are obtained and the validation procedures are presented. It is expected that HERBE will offer qualitative new opportunities for work with fast neutrons at RB reactor including nuclear data determination. (author)

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).

Specific experiments carried out in Germany in order to demonstrate the safety of existing structures

International Nuclear Information System (INIS)

Krutzik, Norbert

2002-01-01

Specific experiments are carried out in Germany in order to demonstrate the safety of existing NPPs. HDR research program includes operational loads testing (pressure test, pressure and temperature test, thermal shock, fatigue); extreme loads (earthquake, aircraft crash, external explosion); internal emergency loads (blowdown, hydrogen combustion, fire, thermal shock, water hammer, condensation loads)

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

International Nuclear Information System (INIS)

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

1994-07-01

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

Proceedings of the specialists meeting on experience with thermal fatigue in LWR piping caused by mixing and stratification

International Nuclear Information System (INIS)

1998-01-01

This specialists meeting on experience with thermal fatigue in LWR piping caused by mixing and stratification, was held in June 1998 in Paris. It included five sessions. Session 1: operating experience (7 papers): Historical perspective; EDF experience with local thermohydraulic phenomena in PWRs: impacts and strategies; Thermal fatigue in safety injection lines of French PWRs: technical problems, regulatory requirements, concerns about other areas; US NRC Regulatory perspective on unanticipated thermal fatigue in LWR piping; Failure to the Residual Heat Removal system suction line pipe in Genkai unit 1 caused by thermal stratification cycling; Emergency Core Cooling System pipe crack incident at Tihange unit 1; Two leakages induced by thermal stratification at the Loviisa power plant). Session 2: thermal hydraulic phenomena (5 papers): Thermal stratification in small pipes with respect to fatigue effects and so called 'Banana effect'; Thermal stratification in the surge line of the Korean next generation reactor; Thermal stratification in horizontal pipes investigated in UPTF-TRAM and HDR facilities; Research on thermal stratification in un-isolable piping of reactor pressure boundary; Thermal mixing phenomena in piping systems: 3D numerical simulation and design considerations. Session 3: response of material and structure (5 papers): Fatigue induced by thermal stratification, Results of tests and calculations of the COUFAST model; Laboratory simulation of thermal fatigue cracking as a basis for verifying life models; Thermo-mechanical analysis methods for the conception and the follow up of components submitted to thermal stratification transients; Piping analysis methods of a PWR surge line for stratified flow; The thermal stratification effect on surge lines, The VVER estimation. Session 4: monitoring aspects (4 papers): Determination of the thermal loadings affecting the auxiliary lines of the reactor coolant system in French PWR plants; Expected and

Tutorial on nuclear thermal propulsion safety for Mars

International Nuclear Information System (INIS)

Buden, D.

1992-01-01

Safety is the prime design requirement for nuclear thermal propulsion (NTP). It must be built in at the initiation of the design process. An understanding of safety concerns is fundamental to the development of nuclear rockets for manned missions to Mars and many other applications that will be enabled or greatly enhanced by the use of nuclear propulsion. To provide an understanding of the basic issues, a tutorial has been prepared. This tutorial covers a range of topics including safety requirements and approaches to meet these requirements, risk and safety analysis methodology, NERVA reliability and safety approach, and life cycle risk assessments

ATLAS program for advanced thermal-hydraulic safety research

International Nuclear Information System (INIS)

Song, Chul-Hwa; Choi, Ki-Yong; Kang, Kyoung-Ho

2015-01-01

Highlights: • Major achievements of the ATLAS program are highlighted in conjunction with both developing advanced light water reactor technologies and enhancing the nuclear safety. • The ATLAS data was shown to be useful for the development and licensing of new reactors and safety analysis codes, and also for nuclear safety enhancement through domestic and international cooperative programs. • A future plan for the ATLAS testing is introduced, covering recently emerging safety issues and some generic thermal-hydraulic concerns. - Abstract: This paper highlights the major achievements of the ATLAS program, which is an integral effect test program for both developing advanced light water reactor technologies and contributing to enhancing nuclear safety. The ATLAS program is closely related with the development of the APR1400 and APR"+ reactors, and the SPACE code, which is a best-estimate system-scale code for a safety analysis of nuclear reactors. The multiple roles of ATLAS testing are emphasized in very close conjunction with the development, licensing, and commercial deployment of these reactors and their safety analysis codes. The role of ATLAS for nuclear safety enhancement is also introduced by taking some examples of its contributions to voluntarily lead to multi-body cooperative programs such as domestic and international standard problems. Finally, a future plan for the utilization of ATLAS testing is introduced, which aims at tackling recently emerging safety issues such as a prolonged station blackout accident and medium-size break LOCA, and some generic thermal-hydraulic concerns as to how to figure out multi-dimensional phenomena and the scaling issue.

Glovebox and Experiment Safety

Science.gov (United States)

Maas, Gerard

2005-12-01

: the identified failure case and the hazardous substance being released (chemical, biological or different).The principle of the glovebox operation is explained, including: mechanical enclosure, air circulation, air filtration and operational modes.Limitations of the glovebox are presented: inability of an experiment fire to be detected by the ASDA, containment only with respect to specified substances, etc. There are requirements induced by the glovebox that the experiment must comply with: Compatibility with the glovebox filter system, thermal limitations, safe without glovebox services, parameter monitoring when a fire hazard is credible, sufficient containment when entering the glovebox and after the experiment, etc.Experiments that are using a glovebox to be operated in shall assess this integrated set-up and the associated operations for compliance to the safety requirements. During this assessment the PSRP shall determine if the provided failure tolerance is sufficient.The gloveboxes that Bradford Engineering (co-) built for human space flight are: USML-1 and 2, MGBX (STS and MIR), MSG, PGBX, LSG-WVA, BGB and PGB. Some of the evolutions are pointed out (experiment services added without compromising safety levels). The major differences of the gloveboxes are presented. For the gloveboxes that are in operation at this time (MSG) or in the near future (BGB, LSG- WVA and PGB) the specific applications are presented.

Thermal-Responsive Polymers for Enhancing Safety of Electrochemical Storage Devices.

Science.gov (United States)

Yang, Hui; Leow, Wan Ru; Chen, Xiaodong

2018-03-01

Thermal runway constitutes the most pressing safety issue in lithium-ion batteries and supercapacitors of large-scale and high-power density due to risks of fire or explosion. However, traditional strategies for averting thermal runaway do not enable the charging-discharging rate to change according to temperature or the original performance to resume when the device is cooled to room temperature. To efficiently control thermal runaway, thermal-responsive polymers provide a feasible and reversible strategy due to their ability to sense and subsequently act according to a predetermined sequence when triggered by heat. Herein, recent research progress on the use of thermal-responsive polymers to enhance the thermal safety of electrochemical storage devices is reviewed. First, a brief discussion is provided on the methods of preventing thermal runaway in electrochemical storage devices. Subsequently, a short review is provided on the different types of thermal-responsive polymers that can efficiently avoid thermal runaway, such as phase change polymers, polymers with sol-gel transitions, and polymers with positive temperature coefficients. The results represent the important development of thermal-responsive polymers toward the prevention of thermal runaway in next-generation smart electrochemical storage devices. © 2018 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Thermal safety characterization on PETN, PBX-9407, LX-10-2, LX-17-1 and detonator in the LLNL's P-ODTX system

Energy Technology Data Exchange (ETDEWEB)

Hsu, P. C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Strout, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Reynolds, J. G. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kahl, E. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ellsworth, F. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Healy, T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2017-09-21

Incidents caused by fire and other thermal events can heat energetic materials that may lead to thermal explosion and result in structural damage and casualty. Thus, it is important to understand the response of energetic materials to thermal insults. The One-Dimensional-Time to Explosion (ODTX) system at the Lawrence Livermore National Laboratory (LLNL) has been used for decades to characterize thermal safety of energetic materials. In this study, an integration of a pressure monitoring element has been added into the ODTX system (P-ODTX) to perform thermal explosion (cook-off) experiments (thermal runaway) on PETN powder, PBX-9407, LX-10-2, LX-17-1, and detonator samples (cup tests). The P-ODTX testing generates useful data (thermal explosion temperature, thermal explosion time, and gas pressures) to assist with the thermal safety assessment of relevant energetic materials and components. This report summarizes the results of P-ODTX experiments that were performed from May 2015 to July 2017. Recent upgrades to the data acquisition system allows for rapid pressure monitoring in microsecond intervals during thermal explosion. These pressure data are also included in the report.

A review on lithium-ion power battery thermal management technologies and thermal safety

Science.gov (United States)

An, Zhoujian; Jia, Li; Ding, Yong; Dang, Chao; Li, Xuejiao

2017-10-01

Lithium-ion power battery has become one of the main power sources for electric vehicles and hybrid electric vehicles because of superior performance compared with other power sources. In order to ensure the safety and improve the performance, the maximum operating temperature and local temperature difference of batteries must be maintained in an appropriate range. The effect of temperature on the capacity fade and aging are simply investigated. The electrode structure, including electrode thickness, particle size and porosity, are analyzed. It is found that all of them have significant influences on the heat generation of battery. Details of various thermal management technologies, namely air based, phase change material based, heat pipe based and liquid based, are discussed and compared from the perspective of improving the external heat dissipation. The selection of different battery thermal management (BTM) technologies should be based on the cooling demand and applications, and liquid cooling is suggested being the most suitable method for large-scale battery pack charged/discharged at higher C-rate and in high-temperature environment. The thermal safety in the respect of propagation and suppression of thermal runaway is analyzed.

A comparison between the IAEA Safety Series 6 thermal environment and a proposed alternative thermal environment

International Nuclear Information System (INIS)

Wix, S.D.; Koski, J.A.

1994-01-01

The present regulations for packaging and transportation of radioactive materials, IAEA Safety Series No. 6; 1985, establish specific criteria for the thermal environment of a hypothetical accident. The regulation states: The scope of this paper is to examine the effects on modeling that result with the Fry proposed thermal boundary conditions. The examination is accomplished by comparing thermal model results using the current IAEA specified thermal environment and the Fry proposed thermal boundary conditions

Research and development program for PWR safety at the CEA reactor thermal hydraulics laboratories

International Nuclear Information System (INIS)

Bernard, M.

1995-01-01

Since the start of the French electronuclear program, the three partners Fermate, EDF and Cea (DRN and IPSN) have devoted considerable effort to research and development for safety issues. In particular an important program on thermal hydraulics was initiated at the beginning of the seventies. It is illustrated by the development of the CATHARE thermalhydraulic safety code which includes physical models derived from a large experimental support program and the construction of the BETHSY integral facility which is aimed to assess both the CATHARE code and the physical relevance of the accident management procedures to be applied on reactors. The state of the art on this program is described with particular emphasis on the capabilities and the assessment of the last version of CATHARE and the lessons drawn from 50 BETHSY tests performed so far. The future plans for safety research cover the following strategy: - to solve the few problems identified on present computing tools and extend the assessment - to solve the few problems identified on present computing tools and extend the assessment - to perform safety studies on the basis of plant operation feedback - to contribute to treating the safety issues related to the future reactors and in particular the case of severe accidents which have to be taken into account from the design stage. The program on severe accidents is aimed to support the design studies performed by the industrial partners and to provide computing tools which model the various phases of severe accidents and will be validated on experiments performed with real and simulating materials. The main lines of the program are: - the development of the TOLBIAC 3D code for the thermal hydraulics of core melt pools, which will be validated against the Bali experiment presently under construction - the Sultan experiment, to study the capability of cooling by external flooding of the reactor vessel - the development of the MC-3D code for core melt

Using the Tritium Plasma Experiment to evaluate ITER PFC safety

International Nuclear Information System (INIS)

Longhurst, G.R.; Anderl, R.A.; Bartlit, J.R.; Causey, R.A.; Haines, J.R.

1993-01-01

The Tritium Plasma Experiment was assembled at Sandia National Laboratories, Livermore to investigate interactions between dense plasmas at low energies and plasma-facing component materials. This apparatus has the unique capability of replicating plasma conditions in a tokamak divertor with particle flux densities of 2 x 10 19 ions/cm 2 · s and a plasma temperature of about 15 eV using a plasma that includes tritium. With the closure of the Tritium Research Laboratory at Livermore, the experiment was moved to the Tritium Systems Test Assembly facility at Los Alamos National Laboratory. An experimental program has been initiated there using the Tritium Plasma Experiment to examine safety issues related to tritium in plasma-facing components, particularly the ITER divertor. Those issues include tritium retention and release characteristics, tritium permeation rates and transient times to coolant streams, surface modification and erosion by the plasma, the effects of thermal loads and cycling, and particulate production. A considerable lack of data exists in these areas for many of the materials, especially beryllium, being considered for use in ITER. Not only will basic material behavior with respect to safety issues in the divertor environment be examined, but innovative techniques for optimizing performance with respect to tritium safety by material modification and process control will be investigated. Supplementary experiments will be carried out at the Idaho National Engineering Laboratory and Sandia National Laboratory to expand and clarify results obtained on the Tritium Plasma Experiment

Operating experience: safety perspective

International Nuclear Information System (INIS)

Piplani, Vivek; Krishnamurthy, P.R.; Kumar, Neeraj; Upadhyay, Devendra

2015-01-01

Operating Experience (OE) provides valuable information for improving NPP safety. This may include events, precursors, deviations, deficiencies, problems, new insights to safety, good practices, lessons and corrective actions. As per INSAG-10, an OE program caters as a fundamental means for enhancing the defence-in-depth at NPPs and hence should be viewed as ‘Continuous Safety Performance Improvement Tool’. The ‘Convention on Nuclear Safety’ also recognizes the OE as a tool of high importance for enhancing the NPP safety and its Article 19 mandates each contracting party to establish an effective OE program at operating NPPs. The lessons drawn from major accidents at Three Mile Island, Chernobyl and Fukushima Daiichi NPPs had prompted nuclear stalwarts to change their safety perspective towards NPPs and to frame sound policies on issues like safety culture, severe accident prevention and mitigation. An effective OE program, besides correcting current/potential problems, help in proactively improving the NPP design, operating and maintenance procedures, practices, training, etc., and thus plays vital role in ensuring safe and efficient operation of NPPs. Further it enhances knowledge with regard to equipment operating characteristics, system performance trends and provides data for quantitative and qualitative safety analysis. Besides all above, an OE program inculcates a learning culture in the organisation and thus helps in continuously enhancing the expertise, technical competency and knowledge base of its staff. Nuclear and Radiation Facilities in India are regulated by Atomic Energy Regulatory Board (AERB). Operating Plants Safety Division (OPSD) of AERB is involved in managing operating experience activities. This paper provides insights about the operating experience program of OPSD, AERB (including its on-line data base namely OPSD STAR) and its utilisation in improving the regulations and safety at Indian NPPs/projects. (author)

Historical perspective of thermal reactor safety in light water reactors

International Nuclear Information System (INIS)

Levy, S.

1986-01-01

A brief history of thermal reactor safety in U.S. light water reactors is provided in this paper. Important shortcomings in safety philosophy evolution versus time are identified and potential corrective actions are suggested. It should be recognized, that this analysis represents only one person's opinion and that most historical accountings reflect the author's biases and specific areas of knowledge. In that sense, many of the examples used in this paper are related to heat transfer and fluid flow safety issues, which explains why it has been included in a Thermal Hydraulics session. One additional note of caution: the value of hindsight and the selective nature of human memory when looking at the past cannot be overemphasized in any historical perspective

Thermal-hydraulic Experiments for Advanced Physical Model Development

International Nuclear Information System (INIS)

Song, Chul Hwa; Baek, W. P.; Yoon, B. J.

2010-04-01

The improvement of prediction models is needed to enhance the safety analysis capability through the fine measurements of local phenomena. To improve the two-phase interfacial area transport model, the various experiments were carried out used SUBO and DOBO. 2x2 and 6x6 rod bundle test facilities were used for the experiment on the droplet behavior. The experiments on the droplet behavior inside a heated rod bundle were focused on the break-up of droplets induced by a spacer grid in a rod bundle geometry. The experiments used GIRLS and JICO and CFD analysis were carried out to comprehend the local condensation of steam jet, turbulent jet induced by condensation and the thermal mixing in a pool. An experimental database of the CHF (Critical Heat Flux) and PDO (Post-dryout) had been constructed. The mechanism of the heat transfer enhancement by surface modifications in nano-fluid was investigated in boiling mode and rapid quenching mode. The special measurement techniques were developed. They are Double -sensor optical void probe, Optic Rod, PIV technique and UBIM system

Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 1

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-07-01

More than 100 papers were presented. The meeting was divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling.

Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 1

International Nuclear Information System (INIS)

2004-01-01

More than 100 papers were presented. The meeting was divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling

Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 2

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-07-01

More than 100 papers presented at the meeting were divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling.

Proceedings of the fourth international topical meeting on nuclear thermal hydraulics, operations and safety. Vol. 2

International Nuclear Information System (INIS)

2004-01-01

More than 100 papers presented at the meeting were divided in 56 sessions and covered the following topics: Plant Operation, Retrofitting and Maintenance Experience; Steam Generator Operation and Maintenance; Artificial Intelligence and Expert Systems; Seismic Technologies for Plant Design and Operations; Aging Management and Life Extension; Two-Phase Flow Modeling and Applications; Severe Accidents and Degraded Core Thermal Hydraulics; Plant Simulators, Analyzers, and Workstations; Advanced Nuclear Fuel Challenges; Recent Nuclear Power Station Decommissioning Experiences in the USA; Application of Probabilistic risk assessment/Probabilistic safety assessment (PRA/PSA) in Design and Modification; Numerical Modeling in Thermal Hydraulics; General Thermal Hydraulics; Severe Accident Management; Licensing and Regulatory Requirements; Advanced Light Water Reactor Designs to Support Reduced Emergency Planning; Best Estimate loss-of-coolant (LOCA) Methodologies; Plant Instrumentation and Control; LWR Fuel Designs for Improved Thermal Hydraulic Performance; Performance Assessment of Radioactive Waste Disposal; Thermal Hydraulics in Passive Reactor Systems; Advances in Man-Machine Interface Design and the Related Human Factors Engineering; Advances in Measurements and Instrumentation; Computer Aided Technology for non-destructive evaluation (NDE) and Plant Maintenance Plant Uprating; Flow-Accelerated Corrosion in Nuclear Power Plants; Advances in Radiological Measurement and Analysis Risk Management and Assessment; Stability in Thermal Hydraulic Systems; Critical heat flux (CHF) and Post Dryout Heat Transfer; Plant Transient and Accident Modeling

Assessment of thermal fatigue crack propagation in safety injection PWR lines

International Nuclear Information System (INIS)

Simos, N.; Reich, M.; Costantino, C.J.; Hartzman, M.

1990-01-01

Cyclic thermal stratification resulting in alternating thermal stresses in pipe cross sections has been identified as the primary cause of high cycle thermal fatigue failure. A number of piping lines in operating plants around the world, susceptible to thermal stratification, have experienced circumferential cracking as a result of high levels of alternating bending stresses. This paper addresses the mechanisms of crack initiation and crack growth and provides estimates of fatigue cycles to failure for a typical safety injection line with such cyclic load history. Utilizing a 3-D finite element analysis, the temperature profile and the corresponding thermal stress field of a complete thermal cycle in a safety injection line consisting of a horizontal pipe section and an elbow, is obtained. Since the observed cracking occurred in the region of the elbow-to-horizontal pipe weld, the analysis performed assessed (1) the impact of the level of local geometric discontinuities on the initiation of an inside surface flaw is greatest and (2) the number of thermal cycles required to drive a small surface crack through the pipe wall. 12 refs., 14 figs., 2 tabs

Approaches to passive safety in advanced thermal reactors

International Nuclear Information System (INIS)

Moses, D.L.

1986-01-01

Since 1980, there has been a proliferation of thermal reactor designs which incorporate passive safety features. The evolution of this trend is briefly traced, and the nature of various passive safety features is discussed with regard to how they have been incorporated into evolving design concepts. The key aspects of the passive safety features include reduced core power density, enhanced passive heat sinks, inherent assured shutdown mechanisms, elimination/minimization of potential leak paths from the primary coolant systems, enhanced robustness of fuel elements and improved coolant chemistry and component materials. An increased reliance on purely passive safety features typically translates into larger reactor structures at reduced power ratings. Proponents of the most innovative concepts seek to offset the increased costs by simplifying licensing requirements and reducing construction time

Improvement of sweating model in 2-Node Model and its application to thermal safety for hot environments

Energy Technology Data Exchange (ETDEWEB)

Ooka, Ryozo [Institute of Industrial Science, The University of Tokyo, 4-6-1 Komaba Meguro-ku, Tokyo 153 8505 (Japan); Minami, Yuriko [Tokyo Electric Power Company, Tokyo (Japan); Sakoi, Tomonori [International Young Researchers Empowerment Center, Shinshu University, Nagano (Japan); Tsuzuki, Kazuyo [National Institute of Advanced Industrial Science and Technology, Tsukuba (Japan); Rijal, H.B. [Integrated Research System for Sustainability Science, The University of Tokyo, Tokyo (Japan)

2010-07-15

Recently, due to global warming and the heat-island effect, more and more people are exposed to the dangers of heat disorders. A hot thermal environment can be evaluated using various indices, such as new Standard Effective Temperature (SET{sup *}) using the 2-Node Model (2 NM), Wet Bulb Globe Temperature (WBGT), Predicted Heat Strain (PHS) model, and so on. The authors aim to develop a safety evaluation approach for hot environments. Subject experiments are performed in a laboratory to comprehend the physiological response of the human body. The results are compared with the computed values from the 2 NM and PHS models, and improved the sweating model in 2 NM in order to take into account the relationship with metabolic rate. A demonstration is provided of using the new sweating model for evaluating thermal safety in a hot environment. (author)

Development of ecologically safety technology for steam-thermal treatment of organic wastes

International Nuclear Information System (INIS)

Juravskij, J.

1997-01-01

The experience on mitigation of the consequences of the Chernobyl's nuclear power station accident proves that the treatment of large amounts of organic and mixed wastes containing radionuclides is a very urgent scientific and technical problem. In this connection a search for new ideas and development of highly efficient and ecologically safety technologies for treatment of organic radioactive wastes has been undertaken. This study is based on use of physico-mechanical properties of various organic materials (wood, rubber-containing composites, plastics, biomass) subjected to thermal decomposition in the overheated water steam medium. Under such conditions, there is a possibility, under relatively low temperatures (400 - 500 deg. C), to realize thermal decomposition and considerably (in 8 - 50 times) to reduce the amount of wastes, to obtain the main concentration of radionuclides in the solid residue and to prevent releases of gaseous products containing radionuclides to the environment. (author). 5 figs, 1 tab

Thermal safety of ultrasound-enhanced ocular drug delivery: A modeling study

Energy Technology Data Exchange (ETDEWEB)

Nabili, Marjan, E-mail: mnabili@gwmail.gwu.edu [Department of Electrical and Computer Engineering, The George Washington University, 800 22nd Street NW, Room 5000, Washington, DC 20052 (United States); Geist, Craig, E-mail: cgeist@mfa.gwu.edu, E-mail: zderic@gwu.edu [Department of Ophthalmology, The George Washington University, 2150 Pennsylvania Avenue NW, Floor 2A, Washington, DC 20037 (United States); Zderic, Vesna, E-mail: cgeist@mfa.gwu.edu, E-mail: zderic@gwu.edu [Department of Biomedical Engineering, The George Washington University, 800 22nd Street NW, Room 6670, Washington, DC 20052 (United States)

2015-10-15

Purpose: Delivery of sufficient amounts of therapeutic drugs into the eye for treatment of various ocular diseases is often a challenging task. Ultrasound was shown to be effective in enhancing ocular drug delivery in the authors’ previous in vitro and in vivo studies. Methods: The study reported here was designed to investigate the safety of ultrasound application and its potential thermal effects in the eye using PZFlex modeling software. The safety limit in this study was set as a temperature increase of no more than 1.5 °C based on regulatory recommendations and previous experimental safety studies. Acoustic and thermal specifications of different human eye tissues were obtained from the published literature. The tissues of particular interest in this modeling safety study were cornea, lens, and the location of optic nerve in the posterior eye. Ultrasound application was modeled at frequencies of 400 kHz–1 MHz, intensities of 0.3–1 W/cm{sup 2}, and exposure duration of 5 min, which were the parameters used in the authors’ previous drug delivery experiments. The baseline eye temperature was 37 °C. Results: The authors’ results showed that the maximal tissue temperatures after 5 min of ultrasound application were 38, 39, 39.5, and 40 °C in the cornea, 39.5, 40, 42, and 43 °C in the center of the lens, and 37.5, 38.5, and 39 °C in the back of the eye (at the optic nerve location) at frequencies of 400, 600, 800 kHz, and 1 MHz, respectively. Conclusions: The ocular temperatures reached at higher frequencies were considered unsafe based on current recommendations. At a frequency of 400 kHz and intensity of 0.8 W/cm{sup 2} (parameters shown in the authors’ previous in vivo studies to be optimal for ocular drug delivery), the temperature increase was small enough to be considered safe inside different ocular tissues. However, the impact of orbital bone and tissue perfusion should be included in future modeling efforts to determine the safety

Thermal safety of ultrasound-enhanced ocular drug delivery: A modeling study

International Nuclear Information System (INIS)

Nabili, Marjan; Geist, Craig; Zderic, Vesna

2015-01-01

Purpose: Delivery of sufficient amounts of therapeutic drugs into the eye for treatment of various ocular diseases is often a challenging task. Ultrasound was shown to be effective in enhancing ocular drug delivery in the authors’ previous in vitro and in vivo studies. Methods: The study reported here was designed to investigate the safety of ultrasound application and its potential thermal effects in the eye using PZFlex modeling software. The safety limit in this study was set as a temperature increase of no more than 1.5 °C based on regulatory recommendations and previous experimental safety studies. Acoustic and thermal specifications of different human eye tissues were obtained from the published literature. The tissues of particular interest in this modeling safety study were cornea, lens, and the location of optic nerve in the posterior eye. Ultrasound application was modeled at frequencies of 400 kHz–1 MHz, intensities of 0.3–1 W/cm 2 , and exposure duration of 5 min, which were the parameters used in the authors’ previous drug delivery experiments. The baseline eye temperature was 37 °C. Results: The authors’ results showed that the maximal tissue temperatures after 5 min of ultrasound application were 38, 39, 39.5, and 40 °C in the cornea, 39.5, 40, 42, and 43 °C in the center of the lens, and 37.5, 38.5, and 39 °C in the back of the eye (at the optic nerve location) at frequencies of 400, 600, 800 kHz, and 1 MHz, respectively. Conclusions: The ocular temperatures reached at higher frequencies were considered unsafe based on current recommendations. At a frequency of 400 kHz and intensity of 0.8 W/cm 2 (parameters shown in the authors’ previous in vivo studies to be optimal for ocular drug delivery), the temperature increase was small enough to be considered safe inside different ocular tissues. However, the impact of orbital bone and tissue perfusion should be included in future modeling efforts to determine the safety of this

Thermal-hydraulics technological strategy roadmap for LWR safety improvement and development

International Nuclear Information System (INIS)

Nakamura, Hideo; Arai, Kenji; Oikawa, Hirohide

2015-01-01

New version of the Thermal-Hydraulics Safety Evaluation Fundamental Technology Enhancement Strategy Roadmap (TH-RM) was developed by the Atomic Energy Society of Japan (AESJ) for LWR safety improvement and development. The 1st version of TH-RM was prepared in 2009 under collaboration of utilities, vendors, universities, research institutes and technical support organizations (TSO) for regulatory body. The revision was made by three sub-working groups (SWGs) by considering the lessons learned from the Fukushima Daiichi Accident. The 'safety assessment' SWG pursued development of computer codes for safety assessment. The 'fundamental technology' SWG pursued safety improvement and risk reduction via accident management (AM) measures by referring the technical map for severe accident (SA) established by the 'severe accident' SWG. Phenomena and components for counter-measures and/or proper prediction are identified by going through SA progression in both reactor and spent-fuel pool of PWR and BWR. Twelve important technology development subjects were identified, which include melt coolability enhancement to maintain integrity of containment vessel. Fact Sheet was developed to describe each of identified and selected R and D subjects. External hazards are also considered how to cope with from thermal-hydraulic safety point of view. This paper summarizes the revised TH-RM with several examples and future perspectives. (author)

GCFR thermal-hydraulic experiments

International Nuclear Information System (INIS)

Schlueter, G.; Baxi, C.B.; Dalle Donne, M.; Gat, U.; Fenech, H.; Hanson, D.; Hudina, M.

1980-01-01

The thermal-hydraulic experimental studies performed and planned for the Gas-Cooled Fast Reactor (GCFR) core assemblies are described. The experiments consist of basic studies performed to obtain correlations, and bundle experiments which provide input for code validation and design verification. These studies have been performed and are planned at European laboratories, US national laboratories, Universities in the US, and at General Atomic Company

The Storage of Thermal Reactor Safety Analysis data (STRESA)

International Nuclear Information System (INIS)

Tanarro Colodron, J.

2016-01-01

Full text: Storage of Thermal Reactor Safety Analysis data (STRESA) is an online information system that contains three technical databases: 1) European Nuclear Research Facilities, open to all online visitors; 2) Nuclear Experiments, available only to registered users; 3) Results Data, being the core content of the information system, its availability depends on the role and organisation of each user. Its main purpose is to facilitate the exchange of experimental data produced by large Euratom funded scientific projects addressing severe accidents, providing at the same time a secure repository for this information. Due to its purpose and architecture, it has become an important asset for networks of excellence as SARNET or NUGENIA. The Severe Accident ResearchNetwork of Excellence (SARNET)was set up in 2004 under the aegis of the research Euratom Framework Programmes to study severe accidents in watercooled nuclear power plants. Coordinated by the IRSN, SARNET unites 43 organizations involved in research on nuclear reactor safety in 18 European countries plus the USA, Canada, South Korea and India. In 2013, SARNET became fully integrated in the Technical Area N2(TA2), named “Severe accidents” of NUGENIA association, devoted to R&D on fission technology of Generation II and III. (author

Development of ecologically safety technology for steam-thermal treatment of organic wastes

Energy Technology Data Exchange (ETDEWEB)

Juravskij, J [Centre for Science, Technology and Industrial Applications, ` ` Tokema` ` , Minsk (Belarus)

1997-02-01

The experience on mitigation of the consequences of the Chernobyl`s nuclear power station accident proves that the treatment of large amounts of organic and mixed wastes containing radionuclides is a very urgent scientific and technical problem. In this connection a search for new ideas and development of highly efficient and ecologically safety technologies for treatment of organic radioactive wastes has been undertaken. This study is based on use of physico-mechanical properties of various organic materials (wood, rubber-containing composites, plastics, biomass) subjected to thermal decomposition in the overheated water steam medium. Under such conditions, there is a possibility, under relatively low temperatures (400 - 500 deg. C), to realize thermal decomposition and considerably (in 8 - 50 times) to reduce the amount of wastes, to obtain the main concentration of radionuclides in the solid residue and to prevent releases of gaseous products containing radionuclides to the environment. (author). 5 figs, 1 tab.

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

Radiological safety design considerations for fusion research experiments

International Nuclear Information System (INIS)

Crase, K.W.; Singh, M.S.

1979-01-01

A wide variety of fusion research experiments are in the planning or construction stages. Two such experiments, the Nova Laser Fusion Facility and the Mirror Fusion Test Facility (MFTF), are currently under construction at Lawrence Livermore Laboratory. Although the plasma chamber vault for MFTF and the Nova target room will have thick concrete walls and roofs, the radiation safety problems are made complex by the numerous requirements for shield wall penetrations. This paper addresses radiation safety considerations for the MFTF and Nova experiments, and the need for integrated safety considerations and safety technology development during the planning stages of fusion experiments

Constitutive model development needs for reactor safety thermal-hydraulic codes

International Nuclear Information System (INIS)

Kelly, J.M.

1998-01-01

This paper discusses the constitutive model development needs for our current and future generation of reactor safety thermal-hydraulic analysis codes. Rather than provide a simple 'shopping list' of models to be improved, a detailed description is given of how a constitutive model works within the computational framework of a current reactor safety code employing the two-fluid model of two-phase flow. The intent is to promote a better understanding of both the types of experiments and the instrumentation needs that will be required in the USNRCs code improvement program. First, a summary is given of the modeling considerations that need to be taken into account when developing constitutive models for use in reactor safety thermal-hydraulic codes. Specifically, the two-phase flow model should be applicable to a control volume formulation employing computational volumes with dimensions on the order of meters but containing embedded structure with a dimension on the order of a centimeter. The closure relations are then required to be suitable when averaged over such large volumes containing millions or even tens of millions of discrete fluid particles (bubbles/drops). This implies a space and time averaging procedure that neglects the intermittency observed in slug and chum turbulent two-phase flows. Furthermore, the geometries encountered in reactor systems are complex, the constitutive relations should therefore be component specific (e.g., interfacial shear in a tube does not represent that in a rod bundle nor in the downcomer). When practicable, future modeling efforts should be directed towards resolving the spatial evolution of two-phase flow patterns through the introduction of interfacial area transport equations and by modeling the individual physical processes responsible for the creation or destruction of interfacial area. Then the example of the implementation and assessment of a subcooled boiling model in a two-fluid code is given. The primary parameter

Safety review of experiments at Albuquerque Operations Office

International Nuclear Information System (INIS)

Elliott, K.

1984-01-01

The Department of Energy (DOE) Albuquerque Operations Office is responsible for the safety overview of nuclear reactor and critical assembly facilities at Sandia National Laboratories, Los Alamos National Laboratory, and the Rocky Flats Plant. The important safety concerns with these facilities involve the complex experiments that are performed, and that is the area emphasized. A determination is made by the Albuquerque Office (AL) with assistance from DOE/OMA whether or not a proposed experiment is an unreviewed safety question. Meetings are held with the contractor to resolve and clarify questions that are generated during the review of the proposed experiment. The AL safety evaluation report is completed and any recommendations are discussed. Prior to the experiment a preoperational appraisal is performed to assure that personnel, procedures, and equipment are in readiness for operations. During the experiment, any abnormal condition is reviewed in detail to determine any safety concerns

Development of thermal-hydraulic models for the safety evaluation of CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Lee, Jae Young; Jung, Yun Sik; Hwang, Gi Suk; Kim, Nam Seok [Handong Univ., Pohang (Korea, Republic of); No, Hee Cheon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

2004-02-15

The objective of the present research is to evaluate the safety analysis for CANDU and to improve the Horizontal Stratification Entrainment Model (HSEM) of RELAP5/MOD3.3. This report includes two items the one is the experimental study of entrainment at horizontal pipe with {+-} 36 .deg. C , {+-} 72 .deg. C branch pies, the other is the model improvement of the moderator heat sink in the Calandria. The off-take experiments on onset of entrainment and branch quality were investigated by using water and air as working fluid, and the experimental data were compared by the previous correlations. The previous correlations could not expect experimental results, thus the weak points of the previous correlations were investigated. The improvement of the previous model continues as the next year research. The thermal hydraulic scaling analysis of SPEL, STERN and ideal linear scaling analysis have been studied. As a result, a new scaling method were needed to design a new experimental facility (HGU). A new scaling method with 1/8 length scale was applied. From these results, the thermal hydraulic model for CFD code simulation was designed and test apparatus has been made. The moderator temperature distribution experiments and CFD code simulation will be continued in next year.

Thermal-Hydraulic Experiment To Test The Stable Operation Of A PIUS Type Reactor

International Nuclear Information System (INIS)

Irianto, Djoko; Kanji, T.; Kukita, Y.

1996-01-01

An advanced type of reaktor concept as the Process Inherent Ultimate Safety (PIUS) reactor was based on intrinsically passive safety considerations. The stable operation of a PIUS type reactor is based on the automation of circulation pump speed. An automatic circulation pump speed control system by using a measurement of the temperature distribution in the lower density lock is proposed the PIUS-type reactor. In principle this control system maintains the fluid temperature at the axial center of the lower density lock at average of the fluid temperatures below and above the lower density lock. This control system will prevent the poison water from penetrating into the core during normal operation. The effectiveness of this control system was successfully confirmed by a series of experiments using atmospheric pressure thermal-hydraulic test loop which simulated the PIUS principle. The experiments such as: start-up and power ramping tests for normal operation simulation and loss of feedwater test for an accident condition simulation, carried out in JAERI

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.

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

Thermal analysis and safety information for metal nanopowders by DSC

Energy Technology Data Exchange (ETDEWEB)

Tseng, J.M.; Huang, S.T. [Institute of Safety and Disaster Prevention Technology, Central Taiwan University of Science and Technology, 666, Buzih Road, Beitun District, Taichung 40601, Taiwan, ROC (China); Duh, Y.S.; Hsieh, T.Y.; Sun, Y.Y. [Department of Safety Health and Environmental Engineering, National United University, Miaoli, Taiwan, ROC (China); Lin, J.Z. [Institute of Safety and Disaster Prevention Technology, Central Taiwan University of Science and Technology, 666, Buzih Road, Beitun District, Taichung 40601, Taiwan, ROC (China); Wu, H.C. [Institute of Occupational Safety and Health, Council of Labor Affairs, Taipei, Taiwan, ROC (China); Kao, C.S., E-mail: jcsk@nuu.edu.tw [Department of Safety Health and Environmental Engineering, National United University, Miaoli, Taiwan, ROC (China)

2013-08-20

Highlights: • Metal nanopowders are common and frequently employed in industry. • Nano iron powder experimental results of T{sub o} were 140–150 °C. • Safety information can benefit relevant metal powders industries. - Abstract: Metal nanopowders are common and frequently employed in industry. Iron is mostly applied in high-performance magnetic materials and pollutants treatment for groundwater. Zinc is widely used in brass, bronze, die casting metal, alloys, rubber, and paints, etc. Nonetheless, some disasters induced by metal powders are due to the lack of related safety information. In this study, we applied differential scanning calorimetry (DSC) and used thermal analysis software to evaluate the related thermal safety information, such as exothermic onset temperature (T{sub o}), peak of temperature (T{sub p}), and heat of reaction (ΔH). The nano iron powder experimental results of T{sub o} were 140–150 °C, 148–158 °C, and 141–149 °C for 15 nm, 35 nm, and 65 nm, respectively. The ΔH was larger than 3900 J/g, 5000 J/g, and 3900 J/g for 15 nm, 35 nm, and 65 nm, respectively. Safety information can benefit the relevant metal powders industries for preventing accidents from occurring.

Small scale thermal-hydraulic experiment for stable operation of a pius-type reactor

International Nuclear Information System (INIS)

Tasaka, K.; Tamaki, M.; Imai, S.; Irianto, I.D.; Tsuji, Y.; Kukita, Y.

1994-01-01

Thermal-hydraulic experiments using a small-scale atmospheric pressure test loop have been performed for the Process Inherent Ultimate Safety (PIUS)-type reactor to develop the new pump speed feedback control system. Three feedback control systems based on the measurement of flow rate, differential pressure, and fluid temperature distribution in the lower density lock have been proposed and confirmed by a series of experiments. Each of the feedback control systems had been verified in the simulation experiment such as a start-up simulation test. The automatic pump speed control based on the fluid temperature at the lower density lock was quite effective to maintain the stratified interface between primary water and borated pool water for stable operation of the reactor. (author)

Sodium Loop Safety Facility W-2 experiment fuel pin rupture detection system

International Nuclear Information System (INIS)

Hoffman, M.A.; Kirchner, T.L.; Meyers, S.C.

1980-05-01

The objective of the Sodium Loop Safety Facility (SLSF) W-2 experiment is to characterize the combined effects of a preconditioned full-length fuel column and slow transient overpower (TOP) conditions on breeder reactor (BR) fuel pin cladding failures. The W-2 experiment will meet this objective by providing data in two technological areas: (1) time and location of cladding failure, and (2) early post-failure test fuel behavior. The test involves a seven pin, prototypic full-length fast test reactor (FTR) fuel pin bundle which will be subjected to a simulated unprotected 5 cents/s reactivity transient overpower event. The outer six pins will provide the necessary prototypic thermal-hydraulic environment for the center pin

The spent fuel safety experiment

International Nuclear Information System (INIS)

Harmms, G.A.; Davis, F.J.; Ford, J.T.

1995-01-01

The Department of Energy is conducting an ongoing investigation of the consequences of taking fuel burnup into account in the design of spent fuel transportation packages. A series of experiments, collectively called the Spent Fuel Safety Experiment (SFSX), has been devised to provide integral benchmarks for testing computer-generated predictions of spent fuel behavior. A set of experiments is planned in which sections of unirradiated fuel rods are interchanged with similar sections of spent PWR fuel rods in a critical assembly. By determining the critical size of the arrays, one can obtain benchmark data for comparison with criticality safety calculations. The integral reactivity worth of the spent fuel can be assessed by comparing the measured delayed critical fuel loading with and without spent fuel. An analytical effort to model the experiments and anticipate the core loadings required to yield the delayed critical conditions runs in parallel with the experimental effort

TMX-U [Tandem Mirror Experiment-Upgrade] tandem-mirror thermal-barrier experiments

International Nuclear Information System (INIS)

Simonen, T.C.; Allen, S.L.; Baldwin, D.E.

1986-01-01

Thermal-barrier experiments have been carried out in the Tandem Mirror Experiment-Upgrade (TMX-U). Measurements of nonambipolar and ambipolar radial transport show that these transport processes, as well as end losses, can be controlled at modest densities and durations. Central-cell heating methods using ion-cyclotron heating (ICH) and neutral-beam injection have been demonstrated. Potential mesurements with recently developed methods indicate that deep thermal barriers can be established

Proceedings of the international meeting on thermal nuclear reactor safety. Vol. 1

International Nuclear Information System (INIS)

1983-02-01

Separate abstracts are included for each of the papers presented concerning current issues in nuclear power plant safety; national programs in nuclear power plant safety; radiological source terms; probabilistic risk assessment methods and techniques; non LOCA and small-break-LOCA transients; safety goals; pressurized thermal shocks; applications of reliability and risk methods to probabilistic risk assessment; human factors and man-machine interface; and data bases and special applications

Proceedings of the international meeting on thermal nuclear reactor safety. Vol. 1

Energy Technology Data Exchange (ETDEWEB)

None

1983-02-01

Separate abstracts are included for each of the papers presented concerning current issues in nuclear power plant safety; national programs in nuclear power plant safety; radiological source terms; probabilistic risk assessment methods and techniques; non LOCA and small-break-LOCA transients; safety goals; pressurized thermal shocks; applications of reliability and risk methods to probabilistic risk assessment; human factors and man-machine interface; and data bases and special applications.

Thermal-hydraulic transient characteristics of ship-propulsion reactor investigated through safety analysis

International Nuclear Information System (INIS)

Fujiki, Kazuo; Asaka, Hideaki; Ishida, Toshihisa

1986-01-01

Thermal-hydraulic behaviors in the reactor of Nuclear Ship ''Mutsu'' were investigated through safety evaluation of operational transients by using RETRAN and COBRA-IV codes. The results were compared to the transient behaviors of typical commercial PWR and the characteristics of transient thermal-hydraulic behaviors in ship-loaded reactor were figured out. ''Mutsu'' reactor has larger thermal margin than commercial PWR because it is designed to be used as ship-propulsion power source in the load-following operation mode. This margin makes transient behavior in general milder than in commercial PWR but high opening pressure set point of main-steam safety valves leads poor heat-sink condition after reactor trip. The effects of other small-sized components are also investigated. The findings in the paper will be helpful in the design of future advanced reactor for nuclear ship. (author)

Ecological safety of thermal power industry and investments

International Nuclear Information System (INIS)

Glebov, V.P.

1995-01-01

Evaluation of ecological safety of domestic fossil fuel thermal power industry is given in comparison with foreign one. Ways of solving ecological problems are considered. They are based on introduction of new technologies, providing decrease of ecological effect, on development of effective ash-and sulfur-trapping, nitrogen purification equipment, on production of ecologically improved fuel. The necessity of investments to power industry is noted

US nuclear safety. Review and experience

International Nuclear Information System (INIS)

Hanauer, S.H.

1977-01-01

The paper deals with the evolution of reactor safety principles, design bases, regulatory requirements, and experience in the United States. Safety concerns have evolved over the years, from reactivity transients and shut-down systems, to blowdowns and containment, to severe design basis accidents and mitigating systems, to the performance of actual materials, systems and humans. The primary safety concerns of one epoch have been superseded in considerable measure by those of later times. Successive plateaus of technical understanding are achieved by solutions being found to earlier problems. Design studies, research, operating experience and regulatory imperatives all contribute to the increased understanding and thus to the safety improvements adopted and accepted. The improvement of safety with time, and the ability of existing reactors to operate safely in the face of new concerns, has confirmed the correctness and usefulness of the defence-in-depth approach and safety margins used in safety design in the United States of America. A regulatory programme such as the one in the United States justifies its great cost by its important contributions to safety. Yet only the designers, constructors and operators of nuclear power plants can actually achieve public safety. The regulatory programme audits, assesses and spot-checks the actual work. Since neither materials nor human beings are flawless, mistakes will be made; that is why defence-in-depth and safety margins are provided. The regulatory programme should enhance safety by decreasing the frequency of uncorrected mistakes. Maintenance of public safety also requires technical and managerial competence and attention in the organizations responsible for nuclear plants as well as regulatory organizations. (author)

Current status and future prospects for thermal-hydraulics and safety research

International Nuclear Information System (INIS)

Park, G.C.

2000-01-01

The present paper is to outline the current activities in Korea for the thermal-hydraulics and safety researches, and furthermore illuminate the future aspect of those field under the umbrella of worldwide nuclear prospect. In Korea, a long-term nuclear research plan has been established since 1992, which was recently funded with a fixed monetary rate of Korean won 1.20 per kWh of electricity produced with nuclear power. 11.5% of the fund is assigned for nuclear safety research in 6 areas. Under this program, 3 axes of research body (KAERI, KINS, University) has been operated with close cooperation. Their role, current activities and long-term plan of each body are introduced in the point of thermal-hydraulics' view. (author)

Lithium-thionyl chloride battery safety

Science.gov (United States)

Carter, B.; Williams, R.; Tsay, F.; Rodriguez, A.; Frank, H.

1982-01-01

Primary lithium cells which use LiAlCl4/SOCl2 electrolyte exhibit high energy density and long life. Currently these cells pose a safety problem since they have been found to vent or explode. This paper summarizes experiments to resolve the safety problem of Li-SOCl2 cells by thermal modeling and identification of possibly hazardous intermediates formed during discharge of these cells. A thermal model and mechanism for the reduction of SOCl2 are presented, as well as a discussion of their application to Li-SOCl2 cell safety.

Lithium-thionyl chloride battery safety

Science.gov (United States)

Carter, B.; Williams, R.; Tsay, F.; Rodriguez, A.; Frank, H.

Primary lithium cells which use LiAlCl4/SOCl2 electrolyte exhibit high energy density and long life. Currently these cells pose a safety problem since they have been found to vent or explode. This paper summarizes experiments to resolve the safety problem of Li-SOCl2 cells by thermal modeling and identification of possibly hazardous intermediates formed during discharge of these cells. A thermal model and mechanism for the reduction of SOCl2 are presented, as well as a discussion of their application to Li-SOCl2 cell safety.

Regulatory support activities of JNES by thermal-hydraulic and safety analyses

International Nuclear Information System (INIS)

Kasahara, Fumio

2008-01-01

Current status and some related topics on regulatory support activities of Japan Nuclear Energy Safety Organization (JNES) by thermal-hydraulic and safety analyses are reported. The safety of nuclear facilities is secured primarily by plant owners and operators. However, the regulatory body NISA (Nuclear and Industrial Safety Agency) has conducted a strict regulation to confirm the adequacy of the site condition as well as the basic and detailed design. The JNES has been conducting independent analyses from applicants (audit analyses, etc.) by direction of NISA and supporting its review. In addition to the audit analysis, thermal-hydraulic and safety analyses are used in such areas as analytical evaluation for investigation of causes of accidents and troubles, level 2 PSA for risk informed regulation, etc. Recent activities of audit analyses are for the application of Tsuruga 3 and 4 (APWR), the spent fuel storage facility for the establishment, and the LMFBR Monju for the core change. For the trouble event evaluation, the criticality accident analysis of Sika1 was carried out and the evaluation of effectiveness of accident management (AM) measure for Tomari 3 (PWR) and Monju was performed. The analytical codes for these evaluations are continuously revised by reflecting the state-of-art technical information and validated using the information provided by the data from JAEA, OECD project, etc. (author)

Energetic Materials Center Report--Small-Scale Safety and Thermal Testing Evaluation of Butyl Nitrate

Energy Technology Data Exchange (ETDEWEB)

Hsu, Peter C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States). Energetic Materials Center; Reynolds, John G. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States). Energetic Materials Center

2013-04-26

Butyl Nitrate (BN) was examined by Small-Scale Safety and Thermal (SSST) Testing techniques to determine its sensitivity to impact, friction, spark and thermal exposure simulating handling and storage conditions. Under the conditions tested, the BN exhibits thermal sensitivity above 150 °C, and does not exhibit sensitive to impact, friction or spark.

Proceedings of the 10th international topical meeting on nuclear thermal hydraulics, operation and safety (NUTHOS-10)

International Nuclear Information System (INIS)

2014-01-01

The 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operations and Safety (NUTHOS-10) in Okinawa, Japan is sponsored by Atomic Energy Society of Japan, in cooperation with the International Atomic Energy Agency, and co-sponsored by American Nuclear Society Thermal Hydraulics Division among others. Enhanced safety and reducing cost are going together, which can be achieved through continued research and development efforts. NUTHOS keeps you abreast of the most updated information in the advancement of science and technology in nuclear thermal hydraulics, operations and safety, and provides you insights into the future. (J.P.N.)

Thermal energy management process experiment

Science.gov (United States)

Ollendorf, S.

1984-01-01

The thermal energy management processes experiment (TEMP) will demonstrate that through the use of two-phase flow technology, thermal systems can be significantly enhanced by increasing heat transport capabilities at reduced power consumption while operating within narrow temperature limits. It has been noted that such phenomena as excess fluid puddling, priming, stratification, and surface tension effects all tend to mask the performance of two-phase flow systems in a 1-g field. The flight experiment approach would be to attack the experiment to an appropriate mounting surface with a 15 to 20 meter effective length and provide a heat input and output station in the form of heaters and a radiator. Using environmental data, the size, location, and orientation of the experiment can be optimized. The approach would be to provide a self-contained panel and mount it to the STEP through a frame. A small electronics package would be developed to interface with the STEP avionics for command and data handling. During the flight, heaters on the evaporator will be exercised to determine performance. Flight data will be evaluated against the ground tests to determine any anomalous behavior.

Sodium Loop Safety Facility W-2 experiment fuel pin rupture detection system. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Hoffman, M.A.; Kirchner, T.L.; Meyers, S.C.

1980-05-01

The objective of the Sodium Loop Safety Facility (SLSF) W-2 experiment is to characterize the combined effects of a preconditioned full-length fuel column and slow transient overpower (TOP) conditions on breeder reactor (BR) fuel pin cladding failures. The W-2 experiment will meet this objective by providing data in two technological areas: (1) time and location of cladding failure, and (2) early post-failure test fuel behavior. The test involves a seven pin, prototypic full-length fast test reactor (FTR) fuel pin bundle which will be subjected to a simulated unprotected 5 cents/s reactivity transient overpower event. The outer six pins will provide the necessary prototypic thermal-hydraulic environment for the center pin.

[Operating Room Nurses' Experiences of Securing for Patient Safety].

Science.gov (United States)

Park, Kwang Ok; Kim, Jong Kyung; Kim, Myoung Sook

2015-10-01

This study was done to evaluate the experience of securing patient safety in hospital operating rooms. Experiential data were collected from 15 operating room nurses through in-depth interviews. The main question was "Could you describe your experience with patient safety in the operating room?". Qualitative data from the field and transcribed notes were analyzed using Strauss and Corbin's grounded theory methodology. The core category of experience with patient safety in the operating room was 'trying to maintain principles of patient safety during high-risk surgical procedures'. The participants used two interactional strategies: 'attempt continuous improvement', 'immersion in operation with sharing issues of patient safety'. The results indicate that the important factors for ensuring the safety of patients in the operating room are manpower, education, and a system for patient safety. Successful and safe surgery requires communication, teamwork and recognition of the importance of patient safety by the surgical team.

Experiments and analyses in support of the US ALMR thermal-hydraulic design

International Nuclear Information System (INIS)

Hunsbedt, A.

1993-01-01

The U.S. Advanced Liquid Metal Reactor (ALMR) which is based on the modular PRISM concept utilizes passive safety characteristics to simplify the reactor design and enhance its safety performance. The relatively small size of each reactor facilitates the use of strong negative feedback with rising temperature for inherent reactivity control and direct, natural air cooling for decay heat removal. The tall, slender reactor geometry of the ALMR enhances uniformity and stability of internal flow distribution during steady state operation and natural circulation flow during transient conditions. The flow uniformity and low operating pressure and temperature of the reactor contributes to high structural margins. A number of experiments and associated analyses have been performed to evaluate natural convection and thermal-hydraulic phenomena experienced under decay heat removal conditions. This paper summarizes these various efforts as described separately below and presents the main results. (author)

Using the Tritium Plasma Experiment to evaluate ITER PFC safety

International Nuclear Information System (INIS)

Longhurst, G.R.; Anderl, R.A.; Bartlit, J.R.; Causey, R.A.; Haines, J.R.

1993-01-01

The Tritium Plasma Experiment was assembled at Sandia National Labs., Livermore and is being moved to the Tritium Systems Test Assembly facility at Los Alamos National Lab. to investigate interactions between dense plasmas at low energies and plasma-facing component materials. This apparatus has the unique capability of replicating plasma conditions in a tokamak divertor with particle flux densities of 2 x 10 23 ions/m 2 .s and a plasma temperature of about 15 eV using a plasma that includes tritium. An experimental program has been initiated using the Tritium Plasma Experiment to examine safety issues related to tritium in plasma-facing components, particularly the ITER divertor. Those issues include tritium retention and release characteristics, tritium permeation rates and transient times to coolant streams, surface modification and erosion by the plasma, the effects of thermal loads and cycling, and particulate production. An industrial consortium led by McDonnell Douglas will design and fabricate the test fixtures

Lithium-thionyl chloride battery safety

Energy Technology Data Exchange (ETDEWEB)

Carter, B.; Frank, H.; Rodriquez, A.; Tsay, F.; Williams, R.

1982-08-01

Primary lithium cells which use LiAlCl/sub 4// SOCl/sub 2/ electrolyte exhibit high energy density and long life. Currently these cells pose a safety problem since they have been found to vent or explode. This paper summarizes experiments to resolve the safety problem of Li-SOCl/sub 2/ cells by thermal modeling and identification of possibly hazardous intermediates formed during discharge of these cells. A thermal model and mechanism for the reduction of SOCl/sub 2/ are presented, as well as a discussion of their application to Li-SOCl/sub 2/ cell safety.

IAEA activities to prepare safety codes and guides for thermal neutron nuclear power plants

International Nuclear Information System (INIS)

Iansiti, E.

1977-01-01

programme is directed towards the safety of thermal nuclear power reactors; however, many of the documents can be very well applied to fast reactors. The first five codes have already completed the procedure and about ten to twelve safety guides will be ready in the course of 1977; the work will proceed then at the rate of about twelve safety guides a year, in the following years. On the basis of the experience gained in the preparation of these documents, and when a substantial part of the documents included in the programme of nuclear power plant safety have been published, it is foreseen that the programme be extended to include other nuclear facilities such as fabrication and reprocessing plants

Preliminary thermal-hydraulic and safety analysis of China DFLL-TBM system

Energy Technology Data Exchange (ETDEWEB)

Li, Wei [School of Nuclear Science and Technology, Xi’an Jiaotong University, No. 28, Xianning West Road, Xi’an, Shanxi 710049 (China); Tian, Wenxi, E-mail: wxtian@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, No. 28, Xianning West Road, Xi’an, Shanxi 710049 (China); Qiu, Suizheng; Su, Guanghui; Jiao, Hong [School of Nuclear Science and Technology, Xi’an Jiaotong University, No. 28, Xianning West Road, Xi’an, Shanxi 710049 (China); Bai, Yunqing; Chen, Hongli [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Wu, Yican, E-mail: yican.Wu@Fds.Org.Cn [Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

2013-06-15

Highlights: • Thermal-hydraulic and safety analysis on DFLL-TBM system is performed. • The TBM FW maximum temperature is 541 °C under steady state condition. • The TBM FW maximum temperature does not exceed the melt point of CLAM steel 1500 °C. • Neither the VV pressurization nor vault pressure build-up goes beyond 0.2 MPa. -- Abstract: China has proposed the dual-functional lithium-lead (DFLL) tritium breeding blanket concept for testing in ITER as a test blanket module (TBM), to demonstrate the technologies of tritium self-sufficiency, high-grade heat extraction and efficient electricity production which are needed for DEMO and fusion power plant. Safety assessment of the TBM and its auxiliary system should be conducted to deal with ITER safety issues directly caused by the TBM system failure during the design process. In this work, three potential initial events (PIEs) – in-vessel loss of helium (He) coolant and ex-vessel loss of He coolant and loss of flow without scram (LOFWS) – were analyzed for the TBM system with a modified version of the RELAP5/MOD3 code containing liquid lithium-lead eutectic (LiPb). The code also comprised an empirical expression for MHD pressure drop relevant to three-dimensional (3D) effect, the Lubarsky–Kaufman convective heat transfer correlation for LiPb flow and the Gnielinski convective heat transfer correlation for He flow. Since both LiPb and He serve as TBM coolants, the LiPb and He ancillary cooling systems were modeled to investigate the thermal-hydraulic characteristic of the TBM system and its influence on ITER safety under those accident conditions. The TBM components and the coolants flow within the TBM were simulated with one-dimensional heat structures and their associated hydrodynamic components. ITER enclosures including vacuum vessel (VV), port cell and TCWS vault were also covered in the model for accident analyses. Through this best estimate approach, the calculation indicated that the current

Environmental safety providing during heat insulation works and using thermal insulation materials

Directory of Open Access Journals (Sweden)

Velichko Evgeny

2017-01-01

Full Text Available This article considers the negative effect of thermal insulating materials and products on human health and environment pollution, particularly in terms of the composition of environmentally hazardous construction products. The authors have analyzed the complex measures for providing ecological safety, sanitary and epidemiological requirements, rules and regulations both during thermal insulation works and throughout the following operation of buildings and premises. The article suggests the protective and preventive measures to reduce and eliminate the negative impact of the proceeding of thermal insulation works on the natural environment and on human health.

US nuclear safety review and experience

International Nuclear Information System (INIS)

Gilinsky, V.

1977-01-01

The nuclear safety review of commercial nuclear power reactors has changed over the years from the relatively simple review of Dresden 1 in 1955 to the highly complex and sophisticated regulatory process which characterizes today's reviews. Four factors have influenced this evolution: (1) maturing of the technology and industry; (2) development of the regulatory process and associated staff; (3) feedback of operating experience; and (4) public awareness and participation. The NRC's safety review responsibilities start before an application is tendered and end when the plant is decommissioned. The safety review for reactor licensing is a comprehensive, two-phase process designed to assure that all the established conservative acceptance criteria are satisfied. Operational safety is assured through a strong inspection and enforcement program which includes shutting down operating facilities when necessary to protect the health and safety of the public. The safety of operating reactors is further insured through close regulation of license changes and selective backfitting of new regulatory requirements. An effective NRC standards development program has been implemented and coordinates closely with the national standards program. A confirmatory safety research program has been developed. Both of these efforts are invaluable to the nuclear safety review because they provide the staff with key tools needed to carry out its regulatory responsibilities. Both have been given increased emphasis since the formation of the NRC in 1975. The safety review process will continue to evolve, but changes will be slower and more deliberate. It will be influenced by standardization, early site reviews and development of advanced reactor concepts. New legislation may make possible changes which will simplify and shorten the regulatory process. Certainly the experience provided by the increasing number and types of operating plants will have a very strong impact on future trends in the

Pressurized-thermal-shock experiments

International Nuclear Information System (INIS)

Whitman, G.D.; McCulloch, R.W.

1982-01-01

The primary objective of the ORNL pressurized-thermal-shock (PTS) experiments is to verify analytical methods that are used to predict the behavior of pressurized-water-reactor vessels under these accident conditions involving combined pressure and thermal loading. The criteria on which the experiments are based are: scale large enough to attain effective flaw border triaxial restraint and a temperature range sufficiently broad to produce a progression from frangible to ductile behavior through the wall at a given time; use of materials that can be completely characterized for analysis; stress states comparable to the actual vessel in zones of potential flaw extension; range of behavior to include cleavage initiation and arrest, cleavage initiation and arrest on the upper shelf, arrest in a high K/sub I/ gradient, warm prestressing, and entirely ductile behavior; long and short flaws with and without stainless steel cladding; and control of loads to prevent vessel burst, except as desired. A PTS test facility is under construction which will enable the establishment and control of wall temperature, cooling rate, and pressure on an intermediate test vessel (ITV) in order to simulate stress states representative of an actual reactor pressure vessel

Program of nuclear criticality safety experiment at JAERI

International Nuclear Information System (INIS)

Kobayashi, Iwao; Tachimori, Shoichi; Takeshita, Isao; Suzaki, Takenori; Ohnishi, Nobuaki

1983-11-01

JAERI is promoting the nuclear criticality safety research program, in which a new facility for criticality safety experiments (Criticality Safety Experimental Facility : CSEF) is to be built for the experiments with solution fuel. One of the experimental researches is to measure, collect and evaluate the experimental data needed for evaluation of criticality safety of the nuclear fuel cycle facilities. Another research area is a study of the phenomena themselves which are incidental to postulated critical accidents. Investigation of the scale and characteristics of the influences caused by the accident is also included in this research. The result of the conceptual design of CSEF is summarized in this report. (author)

Thermal decomposition of nano-enabled thermoplastics: Possible environmental health and safety implications

International Nuclear Information System (INIS)

Sotiriou, Georgios A.; Singh, Dilpreet; Zhang, Fang; Chalbot, Marie-Cecile G.; Spielman-Sun, Eleanor; Hoering, Lutz; Kavouras, Ilias G.; Lowry, Gregory V.; Wohlleben, Wendel; Demokritou, Philip

2016-01-01

Highlights: • Nano-enabled products might reach their end-of-life by thermal decomposition. • Thermal decomposition provides two by-products: released aerosol and residual ash. • Is there any nanofiller release in byproducts? • Risk assessment of potential environmental health implications. - Abstract: Nano-enabled products (NEPs) are currently part of our life prompting for detailed investigation of potential nano-release across their life-cycle. Particularly interesting is their end-of-life thermal decomposition scenario. Here, we examine the thermal decomposition of widely used NEPs, namely thermoplastic nanocomposites, and assess the properties of the byproducts (released aerosol and residual ash) and possible environmental health and safety implications. We focus on establishing a fundamental understanding on the effect of thermal decomposition parameters, such as polymer matrix, nanofiller properties, decomposition temperature, on the properties of byproducts using a recently-developed lab-based experimental integrated platform. Our results indicate that thermoplastic polymer matrix strongly influences size and morphology of released aerosol, while there was minimal but detectable nano-release, especially when inorganic nanofillers were used. The chemical composition of the released aerosol was found not to be strongly influenced by the presence of nanofiller at least for the low, industry-relevant loadings assessed here. Furthermore, the morphology and composition of residual ash was found to be strongly influenced by the presence of nanofiller. The findings presented here on thermal decomposition/incineration of NEPs raise important questions and concerns regarding the potential fate and transport of released engineered nanomaterials in environmental media and potential environmental health and safety implications.

Thermal decomposition of nano-enabled thermoplastics: Possible environmental health and safety implications

Energy Technology Data Exchange (ETDEWEB)

Sotiriou, Georgios A.; Singh, Dilpreet; Zhang, Fang [Center for Nanotechnology and Nanotoxicology, Department of Environmental Health, T.H. Chan School of Public Health, Harvard University, 665 Huntington Ave., Boston, MA 02115 (United States); Chalbot, Marie-Cecile G. [Department of Environmental and Occupational Health, College of Public Health, University of Arkansas for Medical Sciences, Little Rock, AR 72205 (United States); Spielman-Sun, Eleanor [Department of Civil and Environmental Engineering, Carnegie Mellon University, Pittsburgh, PA 15213 (United States); Hoering, Lutz [BASF SE, Material Physics, 67056 Ludwigshafen (Germany); Kavouras, Ilias G. [Department of Environmental and Occupational Health, College of Public Health, University of Arkansas for Medical Sciences, Little Rock, AR 72205 (United States); Lowry, Gregory V. [Department of Civil and Environmental Engineering, Carnegie Mellon University, Pittsburgh, PA 15213 (United States); Wohlleben, Wendel [Center for Nanotechnology and Nanotoxicology, Department of Environmental Health, T.H. Chan School of Public Health, Harvard University, 665 Huntington Ave., Boston, MA 02115 (United States); BASF SE, Material Physics, 67056 Ludwigshafen (Germany); Demokritou, Philip, E-mail: pdemokri@hsph.harvard.edu [Center for Nanotechnology and Nanotoxicology, Department of Environmental Health, T.H. Chan School of Public Health, Harvard University, 665 Huntington Ave., Boston, MA 02115 (United States)

2016-03-15

Highlights: • Nano-enabled products might reach their end-of-life by thermal decomposition. • Thermal decomposition provides two by-products: released aerosol and residual ash. • Is there any nanofiller release in byproducts? • Risk assessment of potential environmental health implications. - Abstract: Nano-enabled products (NEPs) are currently part of our life prompting for detailed investigation of potential nano-release across their life-cycle. Particularly interesting is their end-of-life thermal decomposition scenario. Here, we examine the thermal decomposition of widely used NEPs, namely thermoplastic nanocomposites, and assess the properties of the byproducts (released aerosol and residual ash) and possible environmental health and safety implications. We focus on establishing a fundamental understanding on the effect of thermal decomposition parameters, such as polymer matrix, nanofiller properties, decomposition temperature, on the properties of byproducts using a recently-developed lab-based experimental integrated platform. Our results indicate that thermoplastic polymer matrix strongly influences size and morphology of released aerosol, while there was minimal but detectable nano-release, especially when inorganic nanofillers were used. The chemical composition of the released aerosol was found not to be strongly influenced by the presence of nanofiller at least for the low, industry-relevant loadings assessed here. Furthermore, the morphology and composition of residual ash was found to be strongly influenced by the presence of nanofiller. The findings presented here on thermal decomposition/incineration of NEPs raise important questions and concerns regarding the potential fate and transport of released engineered nanomaterials in environmental media and potential environmental health and safety implications.

ICSBEP-2007, International Criticality Safety Benchmark Experiment Handbook

International Nuclear Information System (INIS)

Blair Briggs, J.

2007-01-01

1 - Description: The Critically Safety Benchmark Evaluation Project (CSBEP) was initiated in October of 1992 by the United Sates Department of Energy. The project quickly became an international effort as scientist from other interested countries became involved. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) is now an official activity of the Organization of Economic Cooperation and Development - Nuclear Energy Agency (OECD-NEA). This handbook contains criticality safety benchmark specifications that have been derived from experiments that were performed at various nuclear critical facilities around the world. The benchmark specifications are intended for use by criticality safety engineers to validate calculational techniques used to establish minimum subcritical margins for operations with fissile material. The example calculations presented do not constitute a validation of the codes or cross section data. The work of the ICSBEP is documented as an International Handbook of Evaluated Criticality Safety Benchmark Experiments. Currently, the handbook spans over 42,000 pages and contains 464 evaluations representing 4,092 critical, near-critical, or subcritical configurations and 21 criticality alarm placement/shielding configurations with multiple dose points for each and 46 configurations that have been categorized as fundamental physics measurements that are relevant to criticality safety applications. The handbook is intended for use by criticality safety analysts to perform necessary validations of their calculational techniques and is expected to be a valuable tool for decades to come. The ICSBEP Handbook is available on DVD. You may request a DVD by completing the DVD Request Form on the internet. Access to the Handbook on the Internet requires a password. You may request a password by completing the Password Request Form. The Web address is: http://icsbep.inel.gov/handbook.shtml 2 - Method of solution: Experiments that are found

Use of safety experience feedback to design new nuclear units

International Nuclear Information System (INIS)

Lange, D.; Crochon, J.P.

1985-06-01

For the designer, and about safety, the experience feedback can take place in 3 fields: the operating experience feedback (incidents analysis), the ''study'' experience feedback (improvement of justification and evolution of safety considerations), and the fabrication experience feedback. Some examples are presented for each field [fr

Operating experience feedback report -- Pressure locking and thermal binding of gate valves

International Nuclear Information System (INIS)

Hsu, C.

1993-03-01

The potential for valve inoperability caused by pressure locking and thermal binding has been known for many years in the nuclear industry. Pressure locking or thermal binding is a common-mode failure mechanism that can prevent a gate valve from opening, and could render redundant trains of safety systems or multiple safety systems inoperable. In spite of numerous generic communications issued in the past by the Nuclear Regulatory Commission (NRC) and industry, pressure locking and thermal binding continues to occur to gate valves installed in safety-related systems of both boding water reactors (BWRs) and pressurized water reactors (PWRs). The generic communications to date have not led to effective industry action to fully identify, evaluate, and correct the problem. This report provides a review of operating events involving these failure mechanisms. As a result of this review this report: (1) identifies conditions when the failure mechanisms have occurred, (2) identifies the spectrum of safety systems that have been subjected to the failure mechanisms, and (3) identifies conditions that may introduce the failure mechanisms under both normal and accident conditions. On the basis of the evaluation of the operating events, the Office for Analysis and Evaluation of Operational Data (AEOD) of the NRC concludes that the binding problems with gate valves are an important safety issue that needs priority NRC and industry attention. This report also provides AEOD's recommendation for actions to effectively prevent the occurrence of valve binding failures

TMX-U thermal-barrier experiments

International Nuclear Information System (INIS)

Simonen, T.C.; Allen, S.L.; Barter, J.D.

1988-01-01

This review of thermal-barrier experiments in the Tandem Mirror Experiment Upgrade (TMX-U) describes our progress at Lawrence Livermore National Laboratory in plasma confinement and central-cell heating. Thermal barriers in TMX-U improved axial confinement by two orders of magnitude over a limited range of densities, compared with confinement in single-cell mirrors at the same ion temperature. Our study shows that central-cell radial nonambipolar confinement scales as neoclassical theory and can be eliminated by floating the end walls. Radial ambipolar losses can also be measured and reduced. The electron energy balance is improved in tandem mirrors to near classical, resulting in T/sub e/ up to 0.28 keV. Electron cyclotron heating (ECH) efficiencies up to 42 percent, with low levels of electron microinstability, were achieved when hot electrons in the thermal barrier were heated to average betas as large as 15 percent. The hot-electron distribution is measured from X rays and is modeled by a Fokker-Planck code that includes heating from cavity radio-frequency (RF) fields. Neutral-beam injection in the central cell created average ion betas up to 5 percent with radial profiles of hot ions that are modeled accurately by a radial Fokker-Planck code. Gas fueling between two fundamental ion cyclotron heating (ICH) resonances resulted in symmetrical heating of passing ions toward both ends

Thin-layer boilover in diesel-oil fires: Determining the increase of thermal hazards and safety distances

International Nuclear Information System (INIS)

Ferrero, Fabio; Munoz, Miguel; Arnaldos, Josep

2007-01-01

A study of the effects of thin-layer boilover on large hydrocarbon fires was carried out. In the experiments, diesel-oil was burned in pools with diameters ranging from 1.5 to 6 m. Previous models used to predict emissive power during the stationary state were analysed and successively modified in order to accurately predict thermal hazard during the water ebullition phase. It was discovered that the increase in emissive power during thin-layer boilover is greater when the pool diameter is smaller. Furthermore, the required increases in safety distances in the case of accidents involving this dangerous phenomenon are provided

LMFBR operational safety: the EBR-II experience

International Nuclear Information System (INIS)

Sackett, J.I.; Allen, N.L.; Dean, E.M.; Fryer, R.M.; Larson, H.A.; Lehto, W.K.

1978-01-01

The mission of the Experimental Breeder Reactor II (EBR-II) has evolved from that of a small LMFBR demonstration plant to a major irradiation-test facility. Because of that evolution, many operational-safety issues have been encountered. The paper describes the EBR-II operational-safety experience in four areas: protection-system design, safety-document preparation, tests of off-normal reactor conditions, and tests of elements with breached cladding

Safety experience on EDF's PWRs

International Nuclear Information System (INIS)

Tanguy, P.

1986-01-01

The french nuclear programme has been widely publicized. In 1985, the total nuclear electricity generated was around 216 GWh, i. e. 70% of the electricity produced by electricity de France (EDF). If we consider only pressurized water reactors, at the end of 1985, 37 units were in operation (32 900 MWe and 5 1300 MWe) and 18 were under construction. I intend to review our experience with the safety of PWR's, but I will first present briefly some aspects related to the safety organization in France and the standardization policy. (author) [pt

IER-297 CED-2: Final Design for Thermal/Epithermal eXperiments with Jemima Plates with Polyethylene and Hafnium

Energy Technology Data Exchange (ETDEWEB)

Nelson, A. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Percher, C. M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Zywiec, W. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Heinrichs, D. P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2018-01-11

This report presents the final design (CED-2) for IER-297, and focuses on 15 critical configurations using highly enriched uranium (HEU) Jemima plates moderated by polyethylene with and without hafnium diluent. The goal of the U.S. Nuclear Criticality Safety Program’s Thermal/Epithermal eXperiments (TEX) is to design and conduct new critical experiments to address high priority nuclear data needs from the nuclear criticality safety and nuclear data communities, with special emphasis on intermediate energy (0.625 eV – 100 keV) assemblies that can be easily modified to include various high priority diluent materials. The TEX (IER 184) CED-1 Report [1], completed in 2012, demonstrated the feasibility of meeting the TEX goals with two existing NCSP fissile assets, plutonium Zero Power Physics Reactor (ZPPR) plates and highly enriched uranium (HEU) Jemima plates. The first set of TEX experiments will focus on using the plutonium ZPPR plates with polyethylene moderator and tantalum diluents.

Structural and Thermal Safety Analysis Report for the Type B Radioactive Waste Transport Package

Energy Technology Data Exchange (ETDEWEB)

Kim, D. H.; Seo, K. S.; Lee, J. C.; Bang, K. S

2007-09-15

We carried out structural safety evaluation for the type B radioactive waste transport package. Requirements for type B packages according to the related regulations such as IAEA Safety Standard Series No. TS-R-1, Korea Most Act. 2001-23 and US 10 CFR Part 71 were evaluated. General requirements for packages such as those for a lifting attachment, a tie-down attachment and pressure condition were considered. For the type B radioactive waste transport package, the structural, thermal and containment analyses were carried out under the normal transport conditions. Also the safety analysis were conducted under the accidental transport conditions. The 9 m drop test, 1 m puncture test, fire test and water immersion test under the accidental transport conditions were consecutively done. The type B radioactive waste transport packages were maintained the structural and thermal integrities.

A review of thermal management and safety for lithium ion batteries

DEFF Research Database (Denmark)

Saeed Madani, Seyed; Swierczynski, Maciej Jozef; Kær, Søren Knudsen

2017-01-01

performance. Therefore,thermal management of batteries is essential for various purposes containing thermal runaway and longstanding of cell functioning period. The favorable outcome of electricdriven vehicles (EDVs) depends on the lithium-ion battery technology. Notwithstanding, the safety concern...... is a considerable technical problem and has become an important factor which might postpones subsequent extension of lithium-ion batteries. This paper reviews different methods for thermal management of lithium-ion batteries. Various methods such as using Phase change materials and using air cooling, straight......Decreasing of fossil fuel sources and ecological worries has spurred global attention in the expansion of developing energy storing systems for electric vehicles (EVs). As a consequence of escalating appeal on new dependable power supplier for hybrid electric vehicles, lithium-ion (Li...

Providing Nuclear Criticality Safety Analysis Education through Benchmark Experiment Evaluation

International Nuclear Information System (INIS)

Bess, John D.; Briggs, J. Blair; Nigg, David W.

2009-01-01

One of the challenges that today's new workforce of nuclear criticality safety engineers face is the opportunity to provide assessment of nuclear systems and establish safety guidelines without having received significant experience or hands-on training prior to graduation. Participation in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and/or the International Reactor Physics Experiment Evaluation Project (IRPhEP) provides students and young professionals the opportunity to gain experience and enhance critical engineering skills.

Large Scale Experiments on Spacecraft Fire Safety

Science.gov (United States)

Urban, David; Ruff, Gary A.; Minster, Olivier; Fernandez-Pello, A. Carlos; Tien, James S.; Torero, Jose L.; Legros, Guillaume; Eigenbrod, Christian; Smirnov, Nickolay; Fujita, Osamu;

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

International handbook of evaluated criticality safety benchmark experiments

International Nuclear Information System (INIS)

2010-01-01

The Criticality Safety Benchmark Evaluation Project (CSBEP) was initiated in October of 1992 by the United States Department of Energy. The project quickly became an international effort as scientists from other interested countries became involved. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) became an official activity of the Organization for Economic Cooperation and Development - Nuclear Energy Agency (OECD-NEA) in 1995. This handbook contains criticality safety benchmark specifications that have been derived from experiments performed at various nuclear critical facilities around the world. The benchmark specifications are intended for use by criticality safety engineers to validate calculational techniques used to establish minimum subcritical margins for operations with fissile material and to determine criticality alarm requirement and placement. Many of the specifications are also useful for nuclear data testing. Example calculations are presented; however, these calculations do not constitute a validation of the codes or cross section data. The evaluated criticality safety benchmark data are given in nine volumes. These volumes span over 55,000 pages and contain 516 evaluations with benchmark specifications for 4,405 critical, near critical, or subcritical configurations, 24 criticality alarm placement / shielding configurations with multiple dose points for each, and 200 configurations that have been categorized as fundamental physics measurements that are relevant to criticality safety applications. Experiments that are found unacceptable for use as criticality safety benchmark experiments are discussed in these evaluations; however, benchmark specifications are not derived for such experiments (in some cases models are provided in an appendix). Approximately 770 experimental configurations are categorized as unacceptable for use as criticality safety benchmark experiments. Additional evaluations are in progress and will be

Thermal responses in underground experiments in a dome salt formation

International Nuclear Information System (INIS)

Llewellyn, G.H.

1977-01-01

To provide design information for a radwaste repository in dome salt, in-situ experiments with nonradioactive heat sources are planned. Three such experiments using electrical heat sources are scheduled to be carried out in a salt dome. The purpose of these experiments is to acquire rock mechanics data to ascertain the structural deformation due to the thermal load imposed, to study brine migration and corrosion, and to provide thermal data. A data acquisition system is provided with these experiments to monitor temperatues, heat fluxes, stresses, and ground displacement. A thermal analysis was made on models of each of these experiments. The objective of the analysis is to verify the capability of making accurate transient temperature predictions by the use of computer modeling techniques. Another purpose is to measure in-situ thermal conductivity and compare the results with measurements taken from core samples. The HEATING5 computer program was used to predict transient temperatures around the experiments for periods up to 2 years using two-dimensional and three-dimensional heat transfer models. The results of analysis are presented with the associated boundary conditions used in the individual models

ABWR (K-6/7) construction experience (computer-based safety system)

International Nuclear Information System (INIS)

Yokomura, T.

1998-01-01

TEPCO applied a digital safety system to Kashiwazaki-Kariwa Nuclear Power Station Unit Nos. 6 and 7, the world's first ABWR plant. Although this was the first time to apply a digital safety logic system in Japan, we were able to complete construction of K-6/7 very successfully and without any delay. TEPCO took a approach of developing a substantial amount of experience in digital non- safety systems before undertaking the design of the safety protection system. This paper describes the history, techniques and experience behind achieving a highly reliable digital safety system. (author)

PNRA Process for Utilizing Experience Feedback for Enhancing Nuclear Safety

International Nuclear Information System (INIS)

Shah, Z.H.

2016-01-01

One of the elements essential for any organization to become a learning organization is to learn from its own and others experience. The importance of utilizing experience feedback for enhancing operational safety is highlighted in nuclear industry again and again and this has resulted in establishment of several national and international forums. In addition, IAEA action plan on nuclear safety issued after Fukushima accident further highlighted the importance of experience sharing among nuclear community to enhance global nuclear safety regime. PNRA utilizes operating experience feedback gathered through different sources in order to improve its regulatory processes. During the review of licensing submissions, special emphasis is given to utilize the lessons learnt from experience feedback relating to nuclear industry within and outside the country. This emphasis has gradually resulted in various safety improvements in the facilities and processes. Accordingly, PNRA has developed a systematic process of evaluation of international operating experience feedback with the aim to create safety conscious approach. This process includes collecting information from different international forums such as IAEA, regulatory bodies of other countries and useful feedback of past accidents followed by its screening, evaluation and suggesting recommendations both for PNRA and its licensees. As a result of this process, several improvements concerning regulatory inspection plans of PNRA as well as in regulatory decision making and operational practices of licensees have been highlighted. This paper will present PNRA approach for utilizing experience feedback in its regulatory processes for enhancing / improving nuclear safety. (author)

Nuclear safety research master plan

Energy Technology Data Exchange (ETDEWEB)

Ha, Jae Joo; Yang, J. U.; Jun, Y. S. and others

2001-06-01

The SRMP (Safety Research Master Plan) is established to cope with the changes of nuclear industry environments. The tech. tree is developed according to the accident progress of the nuclear reactor. The 11 research fields are derived to cover the necessary technologies to ensure the safety of nuclear reactors. Based on the developed tech. tree, the following four main research fields are derived as the main safety research areas: 1. Integrated nuclear safety enhancement, 2. Thermal hydraulic experiment and assessment, 3. Severe accident management and experiment, and 4. The integrity of equipment and structure. The research frame and strategies are also recommended to enhance the efficiency of research activity, and to extend the applicability of research output.

The experiences of research reactor accident to safety improvement

International Nuclear Information System (INIS)

Wiranto, S.

1999-01-01

The safety of reactor operation is the main factor in order that the nuclear technology development program can be held according the expected target. Several experience with research reactor incidents must be learned and understood by the nuclear program personnel, especially for operators and supervisors of RSG-GA. Siwabessy. From the incident experience of research reactor in the world, which mentioned in the book 'Experience with research reactor incidents' by IAEA, 1995, was concluded that the main cause of research reactor accidents is understandless about the safety culture by the nuclear installation personnel. With learn, understand and compare between this experiences and the condition of RSG GA Siwabessy is expended the operators and supervisors more attention about the safety culture, so that RSG GA Siwabessy can be operated successfull, safely according the expected target

HTR-PM Safety requirement and Licensing experience

International Nuclear Information System (INIS)

Li Fu; Zhang Zuoyi; Dong Yujie; Wu Zongxin; Sun Yuliang

2014-01-01

HTR-PM is a 200MWe modular pebble bed high temperature reactor demonstration plant which is being built in Shidao Bay, Weihai, Shandong, China. The main design parameters of HTR-PM were fixed in 2006, the basic design was completed in 2008. The review of Preliminary Safety Analysis Report (PSAR) of HTR-PM was started in April 2008, completed in September 2009. In general, HTR- PM design complies with the current safety requirement for nuclear power plant in China, no special standards are developed for modular HTR. Anyway, Chinese Nuclear Safety Authority, together with the designers, developed some dedicated design criteria for key systems and components and published the guideline for the review of safety analysis report of HTR-PM, based on the experiences from licensing of HTR-10 and new development of nuclear safety. The probabilistic safety goal for HTR-PM was also defined by the safety authority. The review of HTR-PM PSAR lasted for one and a half years, with 3 dialogues meetings and 8 topics meetings, with more than 2000 worksheets and answer sheets. The heavily discussed topics during the PSAR review process included: the requirement for the sub-atmospheric ventilation system, the utilization of PSA in design process, the scope of beyond design basis accidents, the requirement for the qualification of TRISO coating particle fuel, and etc. Because of the characteristics of first of a kind for the demonstration plant, the safety authority emphasized the requirement for the experiment and validation, the PSAR was licensed with certain licensing conditions. The whole licensing process was under control, and was re-evaluated again after Fukushima accident to be shown that the design of HTR-PM complies with current safety requirement. This is a good example for how to license a new reactor. (author)

Comprehensive thermal-hydraulic and thermal-mechanical analysis of core and fuel rods for the safety validation of real refueling at the Kozloduy WWER-440

Energy Technology Data Exchange (ETDEWEB)

Stefanova, S; Panajotov, D; Ilieva, B; Vitkova, M; Simeonova, V; Passage, G; Manolova, M [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika

1996-12-31

Safety analysis aimed at determination of thermal-hydraulic and thermal-mechanical margins of core and fuel rods has been carried out using computer codes COBSOFM and PIN-micro. Thermal-hydraulic calculations for the part of the core with maximum heat flux during steady-state regime show that the coolant, cladding and fuel temperatures are within the design limits. A severe accident with reactor blackout has been simulated. It is found that at 95% probability level there is no boiling crisis anywhere in the core. The thermal-mechanical parameters of working assembly fuel rod with maximum load have been calculated. The assembly linear power reached a maximum of 25 kW/m during the second fuel cycle, the fuel temperature remaining well below 1000{sup o} C. As the fuel assembly with typical power history has enough safety margins, it was proposed to use it for one more cycle. 4 refs., 12 figs.

Experiment on thermal insulation and sodium deposition of shield plug

International Nuclear Information System (INIS)

Hashiguchi, K.; Honda, M.; Shiratori, H.; Ozaki, O.; Suzuki, M.

1986-01-01

A series of experiments on temperature distribution and thermal insulation characteristics was conducted using a reduced scale model of LMFBR shield plug. Observation and measurement of sodium deposition were also conducted on the model after the experiment. The effect of annulus natural convection was clarified for temperature and the thermal insulation characteristics from evaluating the result. Temperature distribution analysis was conducted successfully by combining the general purpose structural analysis program NASTRAN and vertical annulus natural convection analysis program VANAC. Moreover, significant effect was substantiated for the annulus convection barrier to increase the thermal insulation performance, narrow horizontal gap structure to prevent sodium deposition and thermal insulation plates. (author)

Introduction to 'International Handbook of Criticality Safety Benchmark Experiments'

International Nuclear Information System (INIS)

Komuro, Yuichi

1998-01-01

The Criticality Safety Benchmark Evaluation Project (CSBEP) was initiated in 1992 by the United States Department of Energy. The project quickly became an international effort as scientists from other interested countries became involved. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) is now an official activity of the Organization for Economic Cooperation and Development-Nuclear Energy Agency (OECD-NEA). 'International Handbook of Criticality Safety Benchmark Experiments' was prepared and is updated year by year by the working group of the project. This handbook contains criticality safety benchmark specifications that have been derived from experiments that were performed at various nuclear critical facilities around the world. The benchmark specifications are intended for use by criticality safety engineers to validate calculation techniques used. The author briefly introduces the informative handbook and would like to encourage Japanese engineers who are in charge of nuclear criticality safety to use the handbook. (author)

Thermal-hydraulics of the Loviisa reactor pressure vessel overcooling transients

International Nuclear Information System (INIS)

Tuomisto, Harri.

1987-06-01

In the Loviisa reactor pressure vessel safety analyses, the thermal-hydraulics of various overcooling transients has been evaluated to give pertinent initial data for fracture-mechanics calculations. The thermal-hydraulic simulations of the developed overcooling scenarios have been performed using best-estimate thermal-hydraulic computer codes. Experimental programs have been carried out to study phenomena related to natural circulation interruptions in the reactor coolant system. These experiments include buoyancy-induced phenomena such as thermal mixing and stratification of cold high-pressure safety injection water in the cold legs and the downcomer, and oscillations of the single-phase natural circulation. In the probabilistic pressurized thermal shock study, the Loviisa training simulator and the advanced system code RELAP5/MOD2 were utilized to simulate selected sequences. Flow stagnation cases were separately calculated with the REMIX computer program. The methods employed were assessed for these calculations against the plant data and own experiments

VUJE's experience in the field of thermal-hydraulic behaviour of WWER

International Nuclear Information System (INIS)

Klepach, J.

1995-01-01

The thermal-hydraulic behavior (THB) of NPP coolant system and its consequences to nuclear safety of WWER reactors in previous Czechoslovakia has been studied in the VUJE (Nuclear Power Plants Research Institute, Trnava, SK). The institute takes part in the development and verification of its own (SLAP, LENKA, PUMKO, SICHTA, TRACO etc.) and international (DYNAMIKA5) codes for thermal-hydraulic analysis. The verification efforts are concentrated on the WWER specific features such as horizontal steam generators, control and safety system functioning, etc. The whole range of NPP accident analyses is covered by the VUJe staff. The author outlined briefly the WWER specific features as design and implemented improvements in Bohunice V-1 and Mochovce V-1 (WWER 230 model). The pros and cons of the WWER design compared against western type PWR are described. It is believed that although the WWERs are designed under the rules and standards of 1960s, their safety and operational performance can be improved to acceptable level by thorough analysis and appropriate measures. 5 figs

Thermal and Structural Analysis of Beamline Components in the Mu2e Experiment

International Nuclear Information System (INIS)

Martin, Luke Daniel

2016-01-01

Fermi National Accelerator Laboratory will be conducting the high energy particle physics experiment Muons to Electrons (Mu2e). In this experiment, physicists will attempt to witness and understand an ultra-rare process which is the conversion of a muon into the lighter mass electron, without creating additional neutrinos. The experiment is conducted by first generating a proton beam which will be collided into a target within the production solenoid (PS). This creates a high-intensity muon beam which passes through a transport solenoid (TS) and into the detector solenoid (DS). In the detector solenoid the muons will be stopped in an aluminum target and a series of detectors will measure the electrons produced. These components have been named the DS train since they are coupled and travel on a rail system when being inserted or extracted from the DS. To facilitate the installation and removal of the DS train, a set of external stands and a support stand for the instrumentation feed-through bulkhead (IFB) have been designed. Full analysis of safety factors and performance of these two designs has been completed. The detector solenoid itself will need to be maintained to a temperature of 22°C ± 10°C. This will minimize thermal strain and ensure the accurate position of the components is maintained to the tolerance of 2 mm. To reduce the thermal gradient, a passive heating system has been developed and reported.

Preparation of a thermal-hydraulic design method for driver core fuel pins of a new in-pile experimental reactor for FBR safety research

International Nuclear Information System (INIS)

Mizuno, Masahiro; Yamaguchi, Katsuhisa; Uto, Nariaki

1999-07-01

A design study of a new in-pile experimental reactor, SERAPH (Safety Engineering Reactor for Accident PHenomenology), for FBR safety research has progressed at JNC (Japan Nuclear Cycle Development Institute). SERAPH is intended for various in-pile experiments to be performed under quasi-steady state and various transient operation modes. In order to evaluate the driver core performance in conducting such experiments, clarify the relating design issues to be resolved and refine the experimental needs, it is indispensable to comprehend the allowable margin for the thermal-hydraulic fuel pin design since it largely affects the strategy for the driver core design. This report presents a thermal-hydraulic design method for the driver core fuel pins, which is a combination of a two-dimensional time-dependent heat transfer analysis code TAC-2D and a general non-linear finite-element structural analysis code FINAS. In TAC-2D, the allowable spatial mesh and the time step sizes are evaluated. The code is modified so as to treat time-dependent thermal properties, include an improved gap heat-transfer model and treat the change of intra-pin gap width under transient modes, for the purpose of improving the accuracy of evaluating heat transfer characteristics which gives a significant impact on the thermal-hydraulic design. As for FINAS, the number of element nodes and spatial meshes required to obtain adequate accuracy for the thermal stress characteristics of a fuel pellet during transient modes are investigated. In addition, post-processing tools are newly developed to process the calculation results obtained from these codes. The results of this work contribute to advancing the fuel pin design study for SERAPH as well with the investigation on the technique of manufacturing fuel pins. (author)

A HISTORICAL PERSPECTIVE OF NUCLEAR THERMAL HYDRAULICS

Energy Technology Data Exchange (ETDEWEB)

D’Auria, F; Rohatgi, Upendra S.

2017-01-12

The nuclear thermal-hydraulics discipline was developed following the needs for nuclear power plants (NPPs) and, to a more limited extent, research reactors (RR) design and safety. As in all other fields where analytical methods are involved, nuclear thermal-hydraulics took benefit of the development of computers. Thermodynamics, rather than fluid dynamics, is at the basis of the development of nuclear thermal-hydraulics together with the experiments in complex two-phase situations, namely, geometry, high thermal density, and pressure.

Safety engineering experiments of explosives

Energy Technology Data Exchange (ETDEWEB)

Ishikawa, Noboru

1987-07-24

The outline of large scale experiments carried out every year since 1969 to obtain fundamental data and then establish the safety engineering standards concerning the manufacturing, storage and transportation, etc. of all explosives was described. Because it becomes recently difficult to ensure the safety distance in powder magazines and powder plants, the sandwich structure with sand is thought to be suitable as the neighboring barrier walls. The special vertical structure for embankments to provide against a emergency explosion is effective to absorb the blast. Explosion behaviors such as initiating sensitivity, detonation, sympathetic detonation, and shock occurence of the ANFO explosives in place of dynamite and the slurry explosives were studied. The safety engineering standards for the manufacturing and application of explosives were studied to establish because accidents by tabacco fire are not still distinguished. Much data concerning early stage fire fighting, a large quantity of flooding and shock occurence from a assumption of ignition during machining in the propellants manufacturing plant, could be obtained. Basic studies were made to prevent pollution in blasting sites. Collected data are utilized for the safety administration after sufficient discussion. (4 figs, 2 tabs, 3 photos, 17 refs)

Data report for ROSA-IV LSTF gravity-driven safety injection experiment run SB-CL-27

International Nuclear Information System (INIS)

Yonomoto, Taisuke; Saitou, Seishi; Kuroda, Takeshi

1994-03-01

Experimental data are presented for the passive injection test, Run SB-CL-27, conducted at the ROSA-IV Large Scale Test Facility (LSTF) on September 17, 1992. This experiment simulated thermal-hydraulic behavior of a gravity-driven, passive safety injection system during a small-break loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). The injection system consisted of a gravity-driven injection tank, located above the reactor vessel, with connecting lines. The tank was initially filled with water of room temperature at the same pressure as the pressurizer. The connecting lines to the cold leg and to the vessel downcomer were opened at the test initiation. Then, a natural circulation flow developed in the loop which was formed by these lines and the injection tank. The hot water in the cold leg circulated into the upper part of tank and accumulated there causing a significant thermal stratification. This thermal stratification prevented direct-contact condensation of steam from occurring during the subsequent tank drain-down phase. Therefore, no condensation-induced depressurization of the tank, affecting adversely the injection performance, occurred. (author)

Periodic Safety Review of Nuclear Power Plants: Experience of Member States

International Nuclear Information System (INIS)

2010-04-01

Routine reviews of nuclear power plant operation (including modifications to hardware and procedures, operating experience, plant management and personnel competence) and special reviews following major events of safety significance are the primary means of safety verification. In addition, many Member States of the IAEA have initiated systematic safety reassessments, termed periodic safety reviews, of nuclear power plants, to assess the cumulative effects of plant ageing and plant modifications, operating experience, technical developments and siting aspects. The reviews include an assessment of plant design and operation against current safety standards and practices, and they have the objective of ensuring a high level of safety throughout the plant's operating lifetime. They are complementary to the routine and special safety reviews and do not replace them. Periodic safety reviews of nuclear power plants are considered an effective way to obtain an overall view of actual plant safety, and to determine reasonable and practical modifications that should be made in order to maintain a high level of safety. They can be used as a means of identifying time limiting features of the plant in order to determine nuclear power plant operation beyond the designed lifetime. The periodic safety review process can be used to support the decision making process for long term operation or licence renewal. Since 1994, the use of periodic safety reviews by Member States has substantially broadened and confirmed its benefits. Periodic safety review results have, for example, been used by some Member States to help provide a basis for continued operation beyond the current licence term, to communicate more effectively with stakeholders regarding nuclear power plant safety, and to help identify changes to plant operation that enhance safety. This IAEA-TECDOC is intended to assist Member States in the implementation of a periodic safety review. This publication complements the

Analysis of criticality safety of coupled fast-thermal core 'HERBE'

International Nuclear Information System (INIS)

Pesic, M.

1991-01-01

Power excursion during possible fast core flooding is analyzed as serious accident. Model gives short filling time of fast zone with moderator after break of fast core tank. Reactivity increase is determined by computer codes and verified in specific experiments. Measurements of safety rods drop time and reactivity worth are performed. Coupled core kinetics parameters are determined according to model of Avery. Power excursion study, depending on power level threshold and safety instrumentation response time is performed. It was shown that safety system can shut-down reactor safely even in case of highly set power thresholds and partially failure of safety chain. (author)

Study on the thermal ignition of gasoline-air mixture in underground oil depots based on experiment and numerical simulation

Science.gov (United States)

Ou, Yihong; Du, Yang; Jiang, Xingsheng; Wang, Dong; Liang, Jianjun

2010-04-01

The study on the special phenomenon, occurrence process and control mechanism of gasoline-air mixture thermal ignition in underground oil depots is of important academic and applied value for enriching scientific theories of explosion safety, developing protective technology against fire and decreasing the number of fire accidents. In this paper, the research on thermal ignition process of gasoline-air mixture in model underground oil depots tunnel has been carried out by using experiment and numerical simulation methods. The calculation result has been demonstrated by the experiment data. The five stages of thermal ignition course, which are slow oxidation stage, rapid oxidation stage, fire stage, flameout stage and quench stage, have been firstly defined and accurately descried. According to the magnitude order of concentration, the species have been divided into six categories, which lay the foundation for explosion-proof design based on the role of different species. The influence of space scale on thermal ignition in small-scale space has been found, and the mechanism for not easy to fire is that the wall reflection causes the reflux of fluids and changes the distribution of heat and mass, so that the progress of chemical reactions in the whole space are also changed. The novel mathematical model on the basis of unification chemical kinetics and thermodynamics established in this paper provides supplementary means for the analysis of process and mechanism of thermal ignition.

Instrumentation and Control Systems for Sodium thermal hydraulic Experiment Loop for Finned-tube sodium-to-Air heat exchanger (SELFA)

Energy Technology Data Exchange (ETDEWEB)

Kim, Byeong Yeon; Kim, Hyung Mo; Cho, Youn Gil; Kim, Jong Man; Ko, Yung Joo; Kang, Byeong Su; Jung, Min Hwan; Jeong, Ji Young [KAERI, Daejeon (Korea, Republic of)

2016-05-15

A forced-draft sodium-to-air heat exchanger (FHX) is a part of decay heat removal system (DHRS) in Prototype Gen-IV Sodium-cooled fast reactor (PGSFR), which is being developed at Korea Atomic Energy Research Institute (KAERI). Sodium thermal hydraulic Experiment Loop for Finned-tube sodium-to-Air heat exchanger (SELFA) is a test facility for verification and validation of the design code for a forced-draft sodium-to-air heat exchanger (FHX). In this paper, we have provided design and fabrication features for the instrumentation and control systems of SELFA. In general, the instrumentation systems and control systems are coupled for measurement and control of process variables. Instrumentation systems have been designed for investigating thermal-hydraulic characteristics of FHX and control systems have been designed to control the main components (e.g. electromagnetic pumps, heaters, valves etc.) required for test in SELFA. In this paper, we have provided configurations of instrumentation and control systems for Sodium thermal hydraulic Experiment Loop for Finned-tube sodium-to-Air heat exchanger (SELFA). The instrumentation and control systems of SELFA have been implemented based on the expected operation ranges and lesson learned from operational experience of 'Sodium integral effect test loop for safety simulation and assessment-1' (STELLA-1)

Thermal-Hydraulic Experiment Facility (THEF)

International Nuclear Information System (INIS)

Martinell, J.S.

1982-01-01

This paper provides an overview of the Thermal-Hydraulic Experiment Facility (THEF) at the Idaho National Engineering Laboratory (INEL). The overview describes the major test systems, measurements, and data acquisition system, and presents objectives, facility configuration, and results for major experimental projects recently conducted at the THEF. Plans for future projects are also discussed. The THEF is located in the Water Reactor Research Test Facility (WRRTF) area at the INEL

Thermal-hydraulic codes validation for safety analysis of NPPs with RBMK

International Nuclear Information System (INIS)

Brus, N.A.; Ioussoupov, O.E.

2000-01-01

This work is devoted to validation of western thermal-hydraulic codes (RELAP5/MOD3 .2 and ATHLET 1.1 Cycle C) in application to Russian designed light water reactors. Such validation is needed due to features of RBMK reactor design and thermal-hydraulics in comparison with PWR and BWR reactors, for which these codes were developed and validated. These validation studies are concluded with a comparison of calculation results of modeling with the thermal-hydraulics codes with the experiments performed earlier using the thermal-hydraulics test facilities with the experimental data. (authors)

Status and subjects of thermal-hydraulic analysis for next-generation LWRs with passive safety systems

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-03-01

The present status and subjects on thermal-hydraulic analysis for next-generation light water reactors (LWRs) with passive safety systems were summarized based on survey results and discussion by subcommittee on improvement of reactor thermal-hydraulic analysis codes under nuclear code committee in Japan Atomic Energy Research Institute. This survey was performed to promote the research of improvement of reactor thermal-hydraulic analysis codes in future. In the first part of this report, the status and subjects on system analysis and those on evaluation of passive safety system performance are summarized for various types of reactor proposed before. In the second part, the status and subjects on multidimensional two-phase flow analysis are reviewed, since the multidimensional analysis was recognized as one of most important subjects through the investigation in the first part. Besides, databases for bubbly flow and annular dispersed flow were explored, those are needed to assess and verify each multidimensional analytical method. The contents in this report are the forefront of thermal-hydraulic analysis for LWRs and those include current findings for the development of multidimensional two-phase flow analytical method. Thus, we expect that the contents can offer various useful information against the improvement of reactor thermal-hydraulic analysis codes in future. (author)

Electrically Driven Thermal Management: Flight Validation, Experiment Development, Future Technologies

Science.gov (United States)

Didion, Jeffrey R.

2018-01-01

Electrically Driven Thermal Management is an active research and technology development initiative incorporating ISS technology flight demonstrations (STP-H5), development of Microgravity Science Glovebox (MSG) flight experiment, and laboratory-based investigations of electrically based thermal management techniques. The program targets integrated thermal management for future generations of RF electronics and power electronic devices. This presentation reviews four program elements: i.) results from the Electrohydrodynamic (EHD) Long Term Flight Demonstration launched in February 2017 ii.) development of the Electrically Driven Liquid Film Boiling Experiment iii.) two University based research efforts iv.) development of Oscillating Heat Pipe evaluation at Goddard Space Flight Center.

Operational safety experience feedback by means of unusual event reports

International Nuclear Information System (INIS)

1996-07-01

Operational experience of nuclear power plants can be used to great advantage to enhance safety performance provided adequate measures are in place to collect and analyse it and to ensure that the conclusions drawn are acted upon. Feedback of operating experience is thus an extremely important tool to ensure high standards of safety in operational nuclear power plants and to improve the capability to prevent serious accidents and to learn from minor deviations and equipment failures - which can serve as early warnings -to prevent even minor events from occurring. Mechanisms also need to be developed to ensure that operating experience is shared both nationally as well as internationally. The operating experience feedback process needs to be fully and effectively established within the nuclear power plant, the utility, the regulatory organization as well as in other institutions such as technical support organizations and designers. The main purpose of this publication is to reflect the international consensus as to the general principles and practices in the operational safety experience feedback process. The examples of national practices for the whole or for particular parts of the process are given in annexes. The publication complements the IAEA Safety Series No.93 ''Systems for Reporting Unusual Events in Nuclear Power Plants'' (1989) and may also give a general guidance for Member States in fulfilling their obligations stipulated in the Nuclear Safety Convention. Figs, tabs

Operational safety experience feedback by means of unusual event reports

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-07-01

Operational experience of nuclear power plants can be used to great advantage to enhance safety performance provided adequate measures are in place to collect and analyse it and to ensure that the conclusions drawn are acted upon. Feedback of operating experience is thus an extremely important tool to ensure high standards of safety in operational nuclear power plants and to improve the capability to prevent serious accidents and to learn from minor deviations and equipment failures - which can serve as early warnings -to prevent even minor events from occurring. Mechanisms also need to be developed to ensure that operating experience is shared both nationally as well as internationally. The operating experience feedback process needs to be fully and effectively established within the nuclear power plant, the utility, the regulatory organization as well as in other institutions such as technical support organizations and designers. The main purpose of this publication is to reflect the international consensus as to the general principles and practices in the operational safety experience feedback process. The examples of national practices for the whole or for particular parts of the process are given in annexes. The publication complements the IAEA Safety Series No.93 ``Systems for Reporting Unusual Events in Nuclear Power Plants`` (1989) and may also give a general guidance for Member States in fulfilling their obligations stipulated in the Nuclear Safety Convention. Figs, tabs.

A test-type hyper-thermal neutron generator for neutron capture therapy - estimation of neutron energy spectrum by simulation calculations and TOF experiments

International Nuclear Information System (INIS)

Sakurai, Yoshinori; Kobayashi, Tooru; Kobayashi, Katsuhei

1999-01-01

In order to clarify the irradiation characteristics of hyper-thermal neutrons and the feasibility of a hyper-thermal neutron irradiation field for neutron capture therapy, a 'test-type' hyper-thermal neutron generator was designed and made. Graphite of 6 cm thickness and 21 cm diameter was selected as the high temperature scatterer. The scatterer is heated up to 1200 deg. C maximum using molybdenum heaters. The radiation heat is shielded by reflectors of molybdenum and stainless steel. The temperature is measured using three R-type thermo-couples and controlled by a program controller. The total thickness of the generator is designed to be as thin as possible, 20 cm in maximum, in the standing point of the neutron beam intensity. The thermal stability, controllability and safety of the generator at high temperature employment were confirmed by the heating tests. As one of the experiments for the characteristics estimation, the neutron energy spectrum dependent on the scatterer temperature was measured by the TOF (time of flight) method using the LINAC neutron generator. The estimations by simulation calculations were also performed. From the experiment and calculation results, it was confirmed that the neutron temperature shifted higher as the scatterer temperature was higher. The prospect of the feasibility of the 'hyper-thermal neutron irradiation field for NCT' was opened from the estimation results of the generator characteristics by the simulation calculations and experiments

THERMAL EVALUATION OF ALTERNATE SHIPPING CASK FOR GTRI EXPERIMENTS

Energy Technology Data Exchange (ETDEWEB)

Donna Post Guillen

2014-06-01

The Global Threat Reduction Initiative (GTRI) has many experiments yet to be irradiated in support of the High Performance Research Reactor fuels development program. Most of the experiments will be irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL), then later shipped to the Hot Fuel Examination Facility (HFEF) located at the Materials and Fuels Complex for post irradiation examination. To date, the General Electric (GE)-2000 cask has been used to transport GTRI experiments between these facilities. However, the availability of the GE-2000 cask to support future GTRI experiments is at risk. In addition, the internal cavity of the GE-2000 cask is too short to accommodate shipping the larger GTRI experiments. Therefore, an alternate shipping capability is being pursued. The Battelle Energy Alliance, LLC, Research Reactor (BRR) cask has been determined to be the best alternative to the GE-2000 cask. An evaluation of the thermal performance of the BRR cask is necessary before proceeding with fabrication of the newly designed cask hardware and the development of handling, shipping, and transport procedures. This paper presents the results of the thermal evaluation of the BRR cask loaded with a representative set of fueled and non-fueled experiments. When analyzed with identical payloads, experiment temperatures were found to be lower with the BRR cask than with the GE-2000 cask. From a thermal standpoint, the BRR cask was found to be a suitable alternate to the GE-2000 cask.

Thermal and stress analyses of meltdown cups for LMFBR safety experiments using SLSF in-reactor loops

International Nuclear Information System (INIS)

Blomquist, C.A.; Ariman, T.; Pierce, R.D.; Pedersen, D.R.

1977-01-01

A description of a meltdown cup to be used in the SLSF in-reactor experiments is presented. Thermal analyses have shown that the cup is capable of containing and cooling the postulated quantities of molten fuel and steel. The basic loadings for stress analyses were defined and failure modes were determined. It was shown that both the maximum bending stress and maximum tangential stress in the Inconel vessel are below the material yield stress. Additionally, the axial stress in the Inconel vessel was found to be negligible. The shear stress in the wire-formed retaining ring is much below the maximum shear stress. Therefore, the meltdown cup is capable of performing its required function

Progress of thermal hydraulic evaluation methods and experimental studies on a sodium-cooled fast reactor and its safety in Japan

Energy Technology Data Exchange (ETDEWEB)

Kamide, Hideki, E-mail: kamide.hideki@jaea.go.jp; Ohshima, Hiroyuki, E-mail: ohshima.hiroyuki@jaea.go.jp; Sakai, Takaaki, E-mail: sakai.takaaki@jaea.go.jp; Tanaka, Masaaki, E-mail: tanaka.masaaki@jaea.go.jp

2017-02-15

Highlights: • Thermal hydraulic issues for safety design criteria of sodium cooled fast reactors. • Measurement of velocity data in a subchannel surrounded by wire wrapped fuel-pins. • Statistical evaluation of core hot spot temperature during natural circulation. • Simulation of dynamics of molten fuel pool in a core disruptive accident. • V&V procedure of a multi-dimensional thermal hydraulic code on thermal striping. - Abstract: In the framework of the Generation-IV International Forum, the safety design criteria (SDC) incorporating safety-related R&D results on innovative technologies and lessons learned from Fukushima Dai-ichi nuclear power plants accident has been established to provide the set of general criteria for the safety designs of structures, systems and components of Generation-IV Sodium-cooled Fast Reactors (Gen-IV SFRs). A number of thermal-hydraulic evaluations are necessary to meet the concept of the criteria in the design studies of Gen-IV SFRs. This paper focuses on four kinds of thermal-hydraulic issues associated with the SDC, i.e., fuel subassembly thermal-hydraulics, natural circulation decay heat removal, core disruptive accidents, and thermal striping. Progress of evaluation methods on these issues is shown with activities on verification and validation (V&V) and experimental studies towards commercialization of SFR in Japan. These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all possible phenomena in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing our knowledge and technologies down.

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

Fusion reactor safety studies, FY 1977

International Nuclear Information System (INIS)

Darby, J.B. Jr.

1978-04-01

This report reviews the technical progress in the fusion reactor safety studies performed during FY 1977 in the Fusion Power Program at the Argonne National Laboratory. The subjects reported on include safety considerations of the vacuum vessel and first-wall design for the ANL/EPR, the thermal responses of a tokamak reactor first wall, the vacuum wall electrical resistive requirements in relationship to magnet safety, and a major effort is reported on considerations and experiments on air detritiation

Thermal safety analysis of a dry storage cask for the Korean standard spent fuel - 16159

International Nuclear Information System (INIS)

Cha, Jeonghun; Kim, S.N.; Choi, K.W.

2009-01-01

A conceptual dry storage facility, which is based on a commercial dry storage facility, was designed for the Korea standard spent nuclear fuel (SNF) and preliminary thermal safety analysis was performed in this study. To perform the preliminary thermal analysis, a thermal analysis method was proposed. The thermal analysis method consists of 2 parts. By using the method, the surface temperature of the storage canister corresponding to the SNF clad temperature was calculated and the adequate air duct area was decided using the calculation result. The initial temperature of the facility was calculated and the fire condition and half air duct blockage were analyzed. (authors)

Study of system safety evaluation on LTO of national project. Thermal fatigue evaluation of piping systems

International Nuclear Information System (INIS)

Kasahara, Naoto; Itoh, Takamoto; Okazaki, Masakazu; Okuda, Yukihiko; Kamaya, Masayuki; Nakamura, Akira; Nakamura, Hitoshi; Machida, Hideo

2012-01-01

Nuclear piping has various kinds of thermal fatigue failure modes. Main causes of thermal loads are structural responses to fluid temperature changes during plant operation. These phenomena have complex mechanisms and so many patterns, that their problems still occur even though well-known issues. To prevent thermal fatigue due to above thermal loads, the JSME guideline is adopted. Both thermal load and fatigue failure mechanism have been investigated and summarized into the knowledgebase. Numerical simulation methods for thermal fatigue evaluation were studied to replace structural tests. Theses knowledge was utilized to validate and justify the JSME guideline. Furthermore, new studies have been launched to apply above knowledge to enhance plant system safety. (author)

Development of Large-Scale Spacecraft Fire Safety Experiments

DEFF Research Database (Denmark)

Ruff, Gary A.; Urban, David L.; Fernandez-Pello, A. Carlos

2013-01-01

exploration missions outside of low-earth orbit and accordingly, more complex in terms of operations, logistics, and safety. This will increase the challenge of ensuring a fire-safe environment for the crew throughout the mission. Based on our fundamental uncertainty of the behavior of fires in low...... of the spacecraft fire safety risk. The activity of this project is supported by an international topical team of fire experts from other space agencies who conduct research that is integrated into the overall experiment design. The large-scale space flight experiment will be conducted in an Orbital Sciences...

Thermal-hydraulic interfacing code modules for CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Liu, W.S.; Gold, M.; Sills, H. [Ontario Hydro Nuclear, Toronto (Canada)] [and others

1997-07-01

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

Thermal-hydraulic interfacing code modules for CANDU reactors

International Nuclear Information System (INIS)

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

1997-01-01

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

Access safety systems - New concepts from the LHC experience

International Nuclear Information System (INIS)

Ladzinski, T.; Delamare, C.; Luca, S. di; Hakulinen, T.; Hammouti, L.; Havart, F.; Juget, J.F.; Ninin, P.; Nunes, R.; Riesco, T.; Sanchez-Corral Mena, E.; Valentini, F.

2012-01-01

The LHC Access Safety System has introduced a number of new concepts into the domain of personnel protection at CERN. These can be grouped into several categories: organisational, architectural and concerning the end-user experience. By anchoring the project on the solid foundations of the IEC 61508/61511 methodology, the CERN team and its contractors managed to design, develop, test and commission on time a SIL3 safety system. The system uses a successful combination of the latest Siemens redundant safety programmable logic controllers with a traditional relay logic hard wired loop. The external envelope barriers used in the LHC include personnel and material access devices, which are interlocked door-booths introducing increased automation of individual access control, thus removing the strain from the operators. These devices ensure the inviolability of the controlled zones by users not holding the required credentials. To this end they are equipped with personnel presence detectors and the access control includes a state of the art bio-metry check. Building on the LHC experience, new projects targeting the refurbishment of the existing access safety infrastructure in the injector chain have started. This paper summarises the new concepts introduced in the LHC access control and safety systems, discusses the return of experience and outlines the main guiding principles for the renewal stage of the personnel protection systems in the LHC injector chain in a homogeneous manner. (authors)

A review of human thermal comfort experiments in controlled and semi-controlled environments

NARCIS (Netherlands)

Craenendonck, Van Stijn; Lauriks, Leen; Vuye, Cedric; Kampen, Jarl

2018-01-01

There are three main methods to improve thermal comfort in existing buildings: modeling, experiments and measurements. Regarding experiments, no standardized procedure exists. This article provides an answer to the question: “What is the most common practice for human thermal comfort experiments in

Resolution of thermal-hydraulic safety and licensing issues for the system 80+trademark design

International Nuclear Information System (INIS)

Carpentino, S.E.; Ritterbusch, S.E.; Schneider, R.E.

1995-01-01

The System 80+ trademark Standard Design is an evolutionary Advanced Light Water Reactor (ALWR) with a generating capacity of 3931 MWt (1350 MWe). The Final Design Approval (FDA) for this design was issued by the Nuclear Regulatory Commission (NRC) in July 1994. The design certification by the NRC is anticipated by the end of 1995 or early 1996. NRC review of the System 80+ design has involved several new safety issues never before addressed in a regulatory atmosphere. In addition, conformance with the Electric Power Research Institute (EPRI) ALWR Utility Requirements Document (URD) required that the System 80+ plant address nuclear industry concerns with regard to design, construction, operation and maintenance of nuclear power plants. A large number of these issues/concerns deals with previously unresolved generic thermal-hydraulic safety issues and severe accident prevention and mitigation. This paper discusses the thermal-hydraulic analyses and evaluations performed for the System 80+ design to resolve safety and licensing issues relevant to both the Nuclear Stream Supply System (NSSS) and containment designs. For the NSSS design, the Safety Depressurization System mitigation capability and resolution of the boron dilution concern are described. Examples of containment design issues dealing with containment shell strength, robustness of the reactor cavity walls and hydrogen mixing under severe accident conditions are also provided. Finally, the overall approach used in the application of NRC's new (NUREG-1465) radiological source term for System 80+ evaluation is described. The robustness of the System 80+ containment design to withstand severe accident consequences was demonstrated through detailed thermal-hydraulic analyses and evaluations. This advanced design to shown to meet NRC severe accident policy goals and ALWR URD requirements without any special design features and unnecessary costs

Thermal and orbital analysis of Earth monitoring Sun-synchronous space experiments

Science.gov (United States)

Killough, Brian D.

1990-01-01

The fundamentals of an Earth monitoring Sun-synchronous orbit are presented. A Sun-synchronous Orbit Analysis Program (SOAP) was developed to calculate orbital parameters for an entire year. The output from this program provides the required input data for the TRASYS thermal radiation computer code, which in turn computes the infrared, solar and Earth albedo heat fluxes incident on a space experiment. Direct incident heat fluxes can be used as input to a generalized thermal analyzer program to size radiators and predict instrument operating temperatures. The SOAP computer code and its application to the thermal analysis methodology presented, should prove useful to the thermal engineer during the design phases of Earth monitoring Sun-synchronous space experiments.

MCNP benchmark analyses of critical experiments for space nuclear thermal propulsion

International Nuclear Information System (INIS)

Selcow, E.C.; Cerbone, R.J.; Ludewig, H.

1993-01-01

The particle-bed reactor (PBR) system is being developed for use in the Space Nuclear Thermal Propulsion (SNTP) Program. This reactor system is characterized by a highly heterogeneous, compact configuration with many streaming pathways. The neutronics analyses performed for this system must be able to accurately predict reactor criticality, kinetics parameters, material worths at various temperatures, feedback coefficients, and detailed fission power and heating distributions. The latter includes coupled axial, radial, and azimuthal profiles. These responses constitute critical inputs and interfaces with the thermal-hydraulics design and safety analyses of the system

Main lessons for FBR safety study from the CABRI experiments

International Nuclear Information System (INIS)

Sato, Ikken

2006-01-01

CABRI project has been carried out FBR safety study with international cooperation of five nations since 1978. The project consists of four periods such as CABRI-1, CABRI-2, CABRI-FAST and CABRI-RAFT. The objects and the main information for hypothetical core decay accident and safety study of fuel are described. The behavior of core decay accident was studied in the first period (CABRI-1), the safety study of fuel was investigated after the second period (CABRI-2). Change of phenomena at the initial process of core decay accident, comparison between analysis and data of fuel diffusion behavior of neutron horoscope by CABRI-2 experiment, representative in-core experiment under slow TOP conditions, the damaged and undamaged pin under slow TOP conditions, the exterior of CABRI and TREAT and the upper part of CABRI and TREAT reactor are shown. CABRI experiments changed to LWR safety study and a part of TREAT is stated. (S.Y.)

Application of the REMIX thermal mixing calculation program for the Loviisa reactor

International Nuclear Information System (INIS)

Kokkonen, I.; Tuomisto, H.

1987-08-01

The REMIX computer program has been validated to be used in the pressurized thermal shock study of the Loviisa reactor pressure vessel. The program has been verified against the data from the thermal and fluid mixing experiments. These experiments have been carried out in Imatran voima Oy to study thermal mixing of the high-pressure safety injection water in the Loviisa VVER-440 type pressurized water reactor. The verified REMIX-versions were applied to reactor calculations in the probabilistic pressurized thermal shock study of the Loviisa Plant

Halden fuel and material experiments beyond operational and safety limits

International Nuclear Information System (INIS)

Volkov, Boris; Wiesenack, Wolfgang; McGrath, M.; Tverberg, T.

2014-01-01

One of the main tasks of any research reactor is to investigate the behavior of nuclear fuel and materials prior to their introduction into the market. For commercial NPPs, it is important both to test nuclear fuels at a fuel burn-up exceeding current limits and to investigate reactor materials for higher irradiation dose. For fuel vendors such tests enable verification of fuel reliability or for the safety limits to be found under different operational conditions and accident situations. For the latter, in-pile experiments have to be performed beyond some normal limits. The program of fuel tests performed in the Halden reactor is aimed mainly at determining: The thermal FGR threshold, which may limit fuel operational power with burn-up increase, the “lift-off effect” when rod internal pressure exceeds coolant pressure, the effects of high burn-up on fuel behavior under power ramps, fuel relocation under LOCA simulation at higher burn-up, the effect of dry-out on high burn-up fuel rod integrity. This paper reviews some of the experiments performed in the Halden reactor for understanding some of the limits for standard fuel utilization with the aim of contributing to the development of innovative fuels and cladding materials that could be used beyond these limits. (author)

Measuring safety climate in a nuclear power plant - an experience sharing

International Nuclear Information System (INIS)

Vincy, M.U.; Varshney, Aloke; Khot, Pankaj

2016-01-01

In this paper the author discusses the experience gained in safety climate measurement of an Indian nuclear power plant. Safety performance is increasingly part of an organization's sustainable development. Nuclear power stations are falling under the category 'high reliability' industries in the world as far as work safety is concerned. Both the research and the practical experience continually point to two underlying factors that drive safety outcomes: the quality of an organisation's leadership and the resulting culture. After years of development in safety technology and safety management system in the industry, management of nuclear industry world over has come to recognize that safety culture has to be addressed if high standards of health and safety are to be maintained. Therefore, nuclear industries in India have been carrying out measurement of safety climate for more than ten years. The objectives of the study are to examine people's values, attitude, perception, competencies, and patterns of behaviour that determine the commitment to, and effectiveness of health and safety management in the industry based on a questionnaires survey and their analysis

Argonne Bubble Experiment Thermal Model Development III

Energy Technology Data Exchange (ETDEWEB)

Buechler, Cynthia Eileen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2018-01-11

This report describes the continuation of the work reported in “Argonne Bubble Experiment Thermal Model Development” and “Argonne Bubble Experiment Thermal Model Development II”. The experiment was performed at Argonne National Laboratory (ANL) in 2014. A rastered 35 MeV electron beam deposited power in a solution of uranyl sulfate, generating heat and radiolytic gas bubbles. Irradiations were performed at beam power levels between 6 and 15 kW. Solution temperatures were measured by thermocouples, and gas bubble behavior was recorded. The previous report2 described the Monte-Carlo N-Particle (MCNP) calculations and Computational Fluid Dynamics (CFD) analysis performed on the as-built solution vessel geometry. The CFD simulations in the current analysis were performed using Ansys Fluent, Ver. 17.2. The same power profiles determined from MCNP calculations in earlier work were used for the 12 and 15 kW simulations. The primary goal of the current work is to calculate the temperature profiles for the 12 and 15 kW cases using reasonable estimates for the gas generation rate, based on images of the bubbles recorded during the irradiations. Temperature profiles resulting from the CFD calculations are compared to experimental measurements.

Reactor Thermal Hydraulic Numerical Calculation And Modeling

International Nuclear Information System (INIS)

Duong Ngoc Hai; Dang The Ba

2008-01-01

In the paper the results of analysis of thermal hydraulic state models using the numerical codes such as COOLOD, EUREKA and RELAP5 for simulation of the reactor thermal hydraulic states are presented. The calculations, analyses of reactor thermal hydraulic state and safety were implemented using different codes. The received numerical results, which were compared each to other, to experiment measurement of Dalat (Vietnam) research reactor and published results, show their appropriateness and capacity for analyses of different appropriate cases. (author)

NRC staff review of licensee responses to pressure-locking and thermal-binding issue

Energy Technology Data Exchange (ETDEWEB)

Rathbun, H.J.

1996-12-01

Commercial nuclear power plant operating experience has indicated that pressure locking and thermal binding represent potential common mode failure mechanisms that can cause safety-related power-operated gate valves to fail in the closed position, thus rendering redundant safety-related systems incapable of performing their safety functions. In Generic Letter (GL) 95-07, {open_quotes}Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves,{close_quotes} the U.S. Nuclear Regulatory Commission (NRC) staff requested that nuclear power plant licensees take certain actions to ensure that valves susceptible to pressure locking or thermal binding are capable of performing their safety functions within the current licensing bases of the facility. The NRC staff has received summary information from licensees in response to GL 95-07 describing actions they have taken to prevent the occurrence of pressure locking and thermal binding. The NRC staff has developed a systematic process to help ensure uniform and consistent review of licensee submittals in response to GL 95-07.

Thermal Safety Analyses for the Production of Plutonium-238 at the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Hurt, Christopher J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Freels, James D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hobbs, Randy W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jain, Prashant K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Maldonado, G. Ivan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-08-01

There has been a considerable effort over the previous few years to demonstrate and optimize the production of plutonium-238 (²³⁸Pu) at the High Flux Isotope Reactor (HFIR). This effort has involved resources from multiple divisions and facilities at the Oak Ridge National Laboratory (ORNL) to demonstrate the fabrication, irradiation, and chemical processing of targets containing neptunium-237 (²³⁷Np) dioxide (NpO₂)/aluminum (Al) cermet pellets. A critical preliminary step to irradiation at the HFIR is to demonstrate the safety of the target under irradiation via documented experiment safety analyses. The steady-state thermal safety analyses of the target are simulated in a finite element model with the COMSOL Multiphysics code that determines, among other crucial parameters, the limiting maximum temperature in the target. Safety analysis efforts for this model discussed in the present report include: (1) initial modeling of single and reduced-length pellet capsules in order to generate an experimental knowledge base that incorporate initial non-linear contact heat transfer and fission gas equations, (2) modeling efforts for prototypical designs of partially loaded and fully loaded targets using limited available knowledge of fabrication and irradiation characteristics, and (3) the most recent and comprehensive modeling effort of a fully coupled thermo-mechanical approach over the entire fully loaded target domain incorporating burn-up dependent irradiation behavior and measured target and pellet properties, hereafter referred to as the production model. These models are used to conservatively determine several important steady-state parameters including target stresses and temperatures, the limiting condition of which is the maximum temperature with respect to the melting point. The single pellet model results provide a basis for the safety of the irradiations, followed by parametric analyses in the initial prototypical designs

Pressurized-thermal-shock experiments with thick vessels

International Nuclear Information System (INIS)

Bryan, R.H.; Nanstad, R.K.; Merkle, J.G.; Robinson, G.C.; Whitman, G.D.

1986-01-01

Information is provided on the series of pressurized-thermal-shock experiments at the Oak Ridge National Laboratory, motivated by a concern for the behavior of flaws in reactor pressure vessels having welds or shells exhibiting low upper-shelf Charpy impact energies, approx. 68J or less

Analysis of SCARABEE BE+3 experiment with ASTEC-Na and comparison with other SFR safety analysis codes

International Nuclear Information System (INIS)

Bandini, Giacomino; Ederli, Stefano; Perez-Martin, Sara; Pfrang, Werner; Girault, Nathalie; Cloarec, Laure

2017-01-01

The ASTEC-Na code was further developed and assessed in the frame of JASMIN project of the 7th EU Framework Program to extend the original capability of ASTEC, dealing with severe accident analysis in LWR to Sodium-cooled Fast Reactors (SFR). The in-pile BE+3 experiment from the SCARABEE-N program has been simulated with ASTEC-Na for thermal-hydraulic models validation purpose. The adequacy of ASTEC-Na thermal-hydraulic models has been also investigated through the comparison with other safety analysis codes. The analysis of SCARABEE BE+3 test confirms the good performance of ASTEC-Na code in the calculation of single-phase conditions and boiling onset, while larger deviations are encountered in the analysis of the two-phase conditions, mainly regarding the propagation of the boiling front. Furthermore, reasonable agreement was found with other code results. (author)

Preliminary Safety Analysis Report for the Tokamak Physics Experiment

International Nuclear Information System (INIS)

Motloch, C.G.; Bonney, R.F.; Levine, J.D.; Masson, L.S.; Commander, J.C.

1995-04-01

This Preliminary Safety Analysis Report (PSAR), includes an indication of the magnitude of facility hazards, complexity of facility operations, and the stage of the facility life-cycle. It presents the results of safety analyses, safety assurance programs, identified vulnerabilities, compensatory measures, and, in general, the rationale describing why the Tokamak Physics Experiment (TPX) can be safely operated. It discusses application of the graded approach to the TPX safety analysis, including the basis for using Department of Energy (DOE) Order 5480.23 and DOE-STD-3009-94 in the development of the PSAR

Endometrial ablation by rollerball electrocoagulation compared to uterine balloon thermal ablation. Technical and safety aspects.

NARCIS (Netherlands)

Zon-Rabelink, I.A.A. van; Vleugels, M.P.; Merkus, J.M.W.M.; Graaf, R.M. de

2003-01-01

OBJECTIVE: To compare two methods of endometrial ablation, hysteroscopic rollerball electrocoagulation (RBE) and non-hysteroscopic uterine balloon thermal (UBT) ablation (Thermachoice), regarding intra- and post-operative technical complications and safety aspects. STUDY DESIGN: A randomised

The Qualification Experiences for Safety-critical Software of POSAFE-Q

Energy Technology Data Exchange (ETDEWEB)

Kim, Jang Yeol; Son, Kwang Seop; Cheon, Se Woo; Lee, Jang Soo; Kwon, Kee Choon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2009-05-15

Programmable Logic Controllers (PLC) have been applied to the Reactor Protection System (RPS) and the Engineered Safety Feature (ESF)-Component Control System (CCS) as the major safety system components of nuclear power plants. This paper describes experiences on the qualification of the safety-critical software including the pCOS kernel and system tasks related to a safety-grade PLC, i.e. the works done for the Software Verification and Validation, Software Safety Analysis, Software Quality Assurance, and Software Configuration Management etc.

Thermal overload protection for electric motors on safety-related motor-operated valves: Generic Issue II.E.6.1

International Nuclear Information System (INIS)

Rothberg, O.

1988-06-01

NRC regulatory positions, as stated in Regulatory Guide 1.106, Revision 1, have been identified by the Office for Analysis and Evaluation of Operational Data (AEOD) as potential contributors to valve motor burnout. AEOD is particularly concerned about the allowed policy of bypassing thermal overload devices during normal or accident conditions. Regulatory Guide 1.106 favors compromising the function of thermal overload devices in favor of completing the safety-related action of valves. The purpose of this study was to determine if the guidance contained in Regulatory Guide 1.106 is appropriate and, if not, to recommend the necessary changes. This report describes thermal overload devices commonly used to protect safety-related valve operator motors. The regulatory guidelines stated in Regulatory Guide 1.106 along with the limitations of thermal overload protection are discussed. Supplements and alternatives to thermal overload protection are also described. Findings and conclusions of several AEOD reports are discussed. Information obtained from the standard review plan, standard technical specifications, technical specifications from representative plants, and several papers are cited

All-Russia Thermal Engineering Institute experience in using difficult to burn fuels in the power industry

Science.gov (United States)

Tugov, A. N.; Ryabov, G. A.; Shtegman, A. V.; Ryzhii, I. A.; Litun, D. S.

2016-07-01

This article presents the results of the research carried out at the All-Russia Thermal Engineering Institute (VTI) aimed at using saline coal, municipal solid waste and bark waste, sunflower husk, and nesting/ manure materials from poultry farms. The results of saline coal burning experience in Troitsk and Verkhny Tagil thermal power plants (TPP) show that when switching the boiler to this coal, it is necessary to take into account its operating reliability and environmental safety. Due to increased chlorine content in saline coal, the concentration of hydrogen chloride can make over 500 mg/m3. That this very fact causes the sharp increase of acidity in sludge and the resulting damage of hydraulic ash removal system equipment at these power stations has been proven. High concentration of HCl can trigger damage of the steam superheater due to high-temperature corrosion and result in a danger of low-temperature corrosion of air heating surfaces. Besides, increased HCl emissions worsen the environmental characteristics of the boiler operation on the whole. The data on waste-to-energy research for municipal solid waste (MSW) has been generalized. Based on the results of mastering various technologies of MSW thermal processing at special plants nos. 2 and 4 in Moscow, as well as laboratory, bench, and industrial studies, the principal technical solutions to be implemented in the modern domestic thermal power plant with the installed capacity of 24 MW and MSW as the primary fuel type has been developed. The experience of the VTI in burning various kinds of organic waste—bark waste, sunflower husk, and nesting/manure materials from poultry farms—has been analyzed.

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

Preliminary experiments on wastes degradation by thermal plasma

International Nuclear Information System (INIS)

Cota S, G.; Pacheco S, J.; Segovia R, A.; Pena E, R.; Merlo S, L.

1996-01-01

This work presents the fundamental aspects involved in the installation and start up of an experimental equipment for the hazardous wastes degradation using the thermal plasma technology. It is mentioned about the form in which the thermal plasma is generated and the characteristics that its make to be an appropriate technology for the hazardous wastes degradation. Just as the installed structures for to realize the experiments and results of the first studies on degradation, using nylon as problem sample. (Author)

Safety aspects and operating experience of LWR plants in Japan

International Nuclear Information System (INIS)

Aoki, S.; Yoshioka, T.; Toyota, M.; Hinoki, M.

1977-01-01

To develop nuclear power generation for the future, it is necessary to put further emphasis on safety assurance and to endeavour to devise measures to improve plant availability, based on the careful analysis of causes that reduce plant availability. The paper discusses the results of studies on the following items from such viewpoints: (1) Safety and operating experience of LWR nuclear power plants in Japan: operating experience with LWRs; improvements in LWR design during the past ten years; analysis of the factors affecting plant availability; (2) Assurance of safety and measures to increase availability: measures for safety and environmental protection; measures to reduce radiation exposure of employees; appropriateness of maintenance and inspection work; measures to increase plant availability; measures to improve reliability of equipment and components; (3) Future technical problems. (author)

Application of thermal comfort theory in probabilistic safety assessment of a nuclear power plant

International Nuclear Information System (INIS)

Zhou Tao; Sun Canhui; Li Zhenyang; Wang Zenghui

2011-01-01

Human factor errors in probabilistic safety assessment (PSA) of a nuclear power plant (NPP) can be prevented using thermal comfort analysis. In this paper, the THERP + HCR model is modified by using PMV (Predicted Mean Vote) and PPD (Predicted Percentage Dissatisfied) index system, so as to obtain the operator cognitive reliability,and to reflect and analyze human perception, thermal comfort status,and cognitive ability in a specific NPP environment. The mechanism of human factors in the PSA is analyzed by operators of skill, rule and knowledge types. The THERP + HCR model modified by thermal comfort theory can reflect the conditions in actual environment, and optimize reliability analysis of human factors. Improving human thermal comfort for different types of operators reduces adverse factors due to human errors, and provides a safe and optimum decision-making for NPPs. (authors)

Experience of safety and performance improvement for fuel handling equipment

International Nuclear Information System (INIS)

Gyoon Chang, Sang; Hee Lee, Dae

2014-01-01

The purpose of this study is to provide experience of safety and performance improvement of fuel handling equipment for nuclear power plants in Korea. The fuel handling equipment, which is used as an important part of critical processes during the refueling outage, has been improved to enhance safety and to optimize fuel handling procedures. Results of data measured during the fuel reloading are incorporated into design changes. The safety and performance improvement for fuel handling equipment could be achieved by simply modifying the components and improving the interlock system. The experience provided in this study can be useful lessons for further improvement of the fuel handling equipment. (authors)

Coupled neutronics/thermal-hydraulics and safety characteristics of liquid-fueled molten salt reactors

International Nuclear Information System (INIS)

Qiu, Suizheng; Zhang, Dalin; Liu, Minghao; Liu, Limin; Xu, Rongshuan; Gong, Cheng; Su, Guanghui

2016-01-01

Molten salt reactor (MSR) as one candidate of the Generation IV advanced nuclear power systems is attracted more attention in China due to its top ranked fuel cycle and thorium utilization. The MSRs are characterized by using liquid-fuel, which offers complicated coupling problem of neutronics and thermal hydraulics. In this paper, the fundamental model and numerical method are established to calculate and analyze the safety characteristics for liquid-fuel MSRs. The theories and methodologies are applied to the MOSART concept. The liquid-fuel flow effects on neutronics, reactivity coefficients and three operation parameters' influences at steady state are obtained, which provide the basic information for safety analysis. The unprotected loss of flow transient is calculated, the results of which shows the inherent safety characteristics of MOSART due to its strong negative reactivity feedbacks.

Coupled neutronics/thermal-hydraulics and safety characteristics of liquid-fueled molten salt reactors

Energy Technology Data Exchange (ETDEWEB)

Qiu, Suizheng; Zhang, Dalin; Liu, Minghao; Liu, Limin; Xu, Rongshuan; Gong, Cheng; Su, Guanghui [Xi' an Jiaotong Univ. (China). State Key Laboratory of Multiphase Flow in Power Engineering

2016-05-15

Molten salt reactor (MSR) as one candidate of the Generation IV advanced nuclear power systems is attracted more attention in China due to its top ranked fuel cycle and thorium utilization. The MSRs are characterized by using liquid-fuel, which offers complicated coupling problem of neutronics and thermal hydraulics. In this paper, the fundamental model and numerical method are established to calculate and analyze the safety characteristics for liquid-fuel MSRs. The theories and methodologies are applied to the MOSART concept. The liquid-fuel flow effects on neutronics, reactivity coefficients and three operation parameters' influences at steady state are obtained, which provide the basic information for safety analysis. The unprotected loss of flow transient is calculated, the results of which shows the inherent safety characteristics of MOSART due to its strong negative reactivity feedbacks.

Micro-Encapsulated Phase Change Materials: A Review of Encapsulation, Safety and Thermal Characteristics

Directory of Open Access Journals (Sweden)

Ahmed Hassan

2016-10-01

Full Text Available Phase change materials (PCMs have been identified as potential candidates for building energy optimization by increasing the thermal mass of buildings. The increased thermal mass results in a drop in the cooling/heating loads, thus decreasing the energy demand in buildings. However, direct incorporation of PCMs into building elements undermines their structural performance, thereby posing a challenge for building integrity. In order to retain/improve building structural performance, as well as improving energy performance, micro-encapsulated PCMs are integrated into building materials. The integration of microencapsulation PCMs into building materials solves the PCM leakage problem and assures a good bond with building materials to achieve better structural performance. The aim of this article is to identify the optimum micro-encapsulation methods and materials for improving the energy, structural and safety performance of buildings. The article reviews the characteristics of micro-encapsulated PCMs relevant to building integration, focusing on safety rating, structural implications, and energy performance. The article uncovers the optimum combinations of the shell (encapsulant and core (PCM materials along with encapsulation methods by evaluating their merits and demerits.

Plasma thermal energy transport: theory and experiments

International Nuclear Information System (INIS)

Coppi, B.

Experiments on the transport across the magnetic field of electron thermal energy are reviewed (Alcator, Frascati Torus). In order to explain the experimental results, a transport model is described that reconfirmed the need to have an expression for the local diffusion coefficient with a negative exponent of the electron temperature

PGSFR Core Thermal Design Procedure to Evaluate the Safety Margin

Energy Technology Data Exchange (ETDEWEB)

Choi, Sun Rock; Kim, Sang-Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

The Korea Atomic Energy Research Institute (KAERI) has performed a SFR design with the final goal of constructing a prototype plant by 2028. The main objective of the SFR prototype plant is to verify the TRU metal fuel performance, reactor operation, and transmutation ability of high-level wastes. The core thermal design is to ensure the safe fuel performance during the whole plant operation. Compared to the critical heat flux in typical light water reactors, nuclear fuel damage in SFR subassemblies arises from a creep induced failure. The creep limit is evaluated based on the maximum cladding temperature, power, neutron flux, and uncertainties in the design parameters, as shown in Fig. 1. In this work, the core thermal design procedures are compared to verify the present PGSFR methodology based on the nuclear plant design criteria/guidelines and previous SFR thermal design methods. The PGSFR core thermal design procedure is verified based on the nuclear plant design criteria/guidelines and previous methods in LWRs and SFRs. The present method aims to directly evaluate the fuel cladding failure and to assure more safety margin. The 2 uncertainty is similar to 95% one-side tolerance limit of 1.96 in LWRs. The HCFs, ITDP, and MCM reveal similar uncertainty propagation for cladding midwall temperature for typical SFR conditions. The present HCFs are mainly employed from the CRBR except the fuel-related uncertainty such as an incorrect fuel distribution. Preliminary PGSFR specific HCFs will be developed by the end of 2015.

Thermal anchoring of wires in large scale superconducting coil test experiment

International Nuclear Information System (INIS)

Patel, Dipak; Sharma, A.N.; Prasad, Upendra; Khristi, Yohan; Varmora, Pankaj; Doshi, Kalpesh; Pradhan, S.

2013-01-01

Highlights: • We addressed how thermal anchoring in large scale coil test is different compare to small cryogenic apparatus? • We did precise estimation of thermal anchoring length at 77 K and 4.2 K heat sink in large scale superconducting coil test experiment. • We addressed, the quality of anchoring without covering entire wires using Kapton/Teflon tape. • We obtained excellent results in temperature measurement without using GE Varnish by doubling estimated anchoring length. -- Abstract: Effective and precise thermal anchoring of wires in cryogenic experiment is mandatory to measure temperature in milikelvin accuracy and to avoid unnecessary cooling power due to additional heat conduction from room temperature (RT) to operating temperature (OT) through potential, field, displacement and stress measurement instrumentation wires. Instrumentation wires used in large scale superconducting coil test experiments are different compare to cryogenic apparatus in terms of unique construction and overall diameter/area due to errorless measurement in large time-varying magnetic field compare to small cryogenic apparatus, often shielded wires are used. Hence, along with other variables, anchoring techniques and required thermal anchoring length are entirely different in this experiment compare to cryogenic apparatus. In present paper, estimation of thermal anchoring length of five different types of instrumentation wires used in coils test campaign at Institute for Plasma Research (IPR), India has been discussed and some temperature measurement results of coils test campaign have been presented

Experience with thermal recycle of plutonium and uranium

International Nuclear Information System (INIS)

Beer, O.; Schlosser, G.; Spielvogel, F.

1985-01-01

The Federal Republic of Germany (FRG) decided to close the fuel cycle by erecting the reprocessing plant WA350 at Wackersdorf. As long as the plutonium supply from reprocessing plants exceeds the plutonium demand of fast breeder reactors, recycling of plutonium in LWR's is a convenient solution by which a significant advanced uranium utilization is achieved. The demonstration of plutonium recycling performed to date in the FRG in BWR's and PWR's shows that thermal plutonium recycling on an industrial scale is feasible and that the usual levels of reliability and safety can be achieved in reactor operation. The recycling of reprocessed uranium is presently demonstrated in the FRG, too. As regards fuel cycle economy thermal recycling allows savings in natural uranium and separative work. Already under present cost conditions the fuel cycle costs for mixed oxide or enriched reprocessed uranium fuel assemblies are equal or even lower than for usual uranium fuel assemblies

Feedback of safety - related operational experience: Lessons learned

Energy Technology Data Exchange (ETDEWEB)

Elias, D [Commonwealth Edison Co. (United States)

1997-09-01

The presentation considers the following aspects of feedback of safety-related operational experience: lessons learned program, objectives, personnel characteristics; three types of documents for transmitting lessons learned issues.

Feedback of safety - related operational experience: Lessons learned

International Nuclear Information System (INIS)

Elias, D.

1997-01-01

The presentation considers the following aspects of feedback of safety-related operational experience: lessons learned program, objectives, personnel characteristics; three types of documents for transmitting lessons learned issues

Experiments and numerical simulations of fluctuating thermal stratification in a branch pipe

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Akira; Murase, Michio; Sasaki, Toru [Inst. of Nuclear Safety System Inc., Mihama, Fukui (Japan); Takenaka, Nobuyuki; Hamatani, Daisuke [Kobe Univ. (Japan)

2002-09-01

Many pipes branch off from the main pipe in plants. When the main flow in the main pipe is hotter than a branch pipe that branches off downward, the hot water penetrates into the branch pipe with the cavity flow that is induced by the main flow and causes thermal stratification. If the interface of the stratification fluctuates in an occluded branch pipe, thermal fatigue may occur in pipe wall. Some experiments and numerical simulations were conducted to elucidate the mechanism of this fluctuating thermal stratification. The vortex structures were observed in the experiments of straight or bent branch pipes. When the main flow was heated and the thermal stratification interface was at the elbow, a ''burst'' phenomenon occurred in the interface in connection with large heat fluctuation. The effects of pipe shape on the length of penetration were investigated in order to modify simulation conditions. The vortex structures and the fluctuating thermal stratification at elbow in the numerical simulation showed good agreement with experiments. (author)

Resolution of thermal-hydraulic safety and licensing issues for the system 80+{sup {trademark}} design

Energy Technology Data Exchange (ETDEWEB)

Carpentino, S.E.; Ritterbusch, S.E.; Schneider, R.E. [ABB-Combustion Engineering, Windsor, CT (United States)] [and others

1995-09-01

The System 80+{sup {trademark}} Standard Design is an evolutionary Advanced Light Water Reactor (ALWR) with a generating capacity of 3931 MWt (1350 MWe). The Final Design Approval (FDA) for this design was issued by the Nuclear Regulatory Commission (NRC) in July 1994. The design certification by the NRC is anticipated by the end of 1995 or early 1996. NRC review of the System 80+ design has involved several new safety issues never before addressed in a regulatory atmosphere. In addition, conformance with the Electric Power Research Institute (EPRI) ALWR Utility Requirements Document (URD) required that the System 80+ plant address nuclear industry concerns with regard to design, construction, operation and maintenance of nuclear power plants. A large number of these issues/concerns deals with previously unresolved generic thermal-hydraulic safety issues and severe accident prevention and mitigation. This paper discusses the thermal-hydraulic analyses and evaluations performed for the System 80+ design to resolve safety and licensing issues relevant to both the Nuclear Stream Supply System (NSSS) and containment designs. For the NSSS design, the Safety Depressurization System mitigation capability and resolution of the boron dilution concern are described. Examples of containment design issues dealing with containment shell strength, robustness of the reactor cavity walls and hydrogen mixing under severe accident conditions are also provided. Finally, the overall approach used in the application of NRC`s new (NUREG-1465) radiological source term for System 80+ evaluation is described. The robustness of the System 80+ containment design to withstand severe accident consequences was demonstrated through detailed thermal-hydraulic analyses and evaluations. This advanced design to shown to meet NRC severe accident policy goals and ALWR URD requirements without any special design features and unnecessary costs.

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)

Application of a statistical thermal design procedure to evaluate the PWR DNBR safety analysis limits

International Nuclear Information System (INIS)

Robeyns, J.; Parmentier, F.; Peeters, G.

2001-01-01

In the framework of safety analysis for the Belgian nuclear power plants and for the reload compatibility studies, Tractebel Energy Engineering (TEE) has developed, to define a 95/95 DNBR criterion, a statistical thermal design method based on the analytical full statistical approach: the Statistical Thermal Design Procedure (STDP). In that methodology, each DNBR value in the core assemblies is calculated with an adapted CHF (Critical Heat Flux) correlation implemented in the sub-channel code Cobra for core thermal hydraulic analysis. The uncertainties of the correlation are represented by the statistical parameters calculated from an experimental database. The main objective of a sub-channel analysis is to prove that in all class 1 and class 2 situations, the minimum DNBR (Departure from Nucleate Boiling Ratio) remains higher than the Safety Analysis Limit (SAL). The SAL value is calculated from the Statistical Design Limit (SDL) value adjusted with some penalties and deterministic factors. The search of a realistic value for the SDL is the objective of the statistical thermal design methods. In this report, we apply a full statistical approach to define the DNBR criterion or SDL (Statistical Design Limit) with the strict observance of the design criteria defined in the Standard Review Plan. The same statistical approach is used to define the expected number of rods experiencing DNB. (author)

On dependence of thermal transport on the safety factor q in tokamaks

International Nuclear Information System (INIS)

Hirose, A.; Singh, A.; Livingstone, S.; Liu, D.

2005-01-01

Mixing length estimates of the anomalous ion and electron thermal diffusivities caused by the ITG and ETG mode, respectively, have revealed that both diffusivities are proportional to the safety factor q. In the case of ITG mode, the maximum χ i occurs at long wavelengths where coupling to the ion acoustic mode is dominant while ETG driven χ e peaks at wavelengths comparable to the electron skin depth. (author)

Recent results from TMX-U thermal barrier experiments

International Nuclear Information System (INIS)

Molvik, A.W.; Allen, S.; Barter, J.

1984-01-01

The Tandem Mirror Experiment-Upgrade (TMX-U) device was designed to study plasma confinement in a tandem mirror with thermal barriers. Previously the author reported improved axial confinement with high end-plug potentials, consistent with thermal barrier operation. Now, the existence of thermal barriers in TMX-U confirmed by measuring the axial potential profile. Specifically, measured the change in energy of a 5-keV deuterium neutral beam that is injected nearly parallel to the axis and is ionized between the barrier and the central cell. The authors found that the barrier potential is lower than the central cell potential, as required for a thermal barrier. The peak potential is at least 2.4 keV, as determined from the minimum energy of end loss ions. In addition, radial transport is reduced by the use of floating and electrodes that map to concentric cylinders in the central cell. Sloshing ions continue to be microstable

Bootstrap and Order Statistics for Quantifying Thermal-Hydraulic Code Uncertainties in the Estimation of Safety Margins

Directory of Open Access Journals (Sweden)

Enrico Zio

2008-01-01

Full Text Available In the present work, the uncertainties affecting the safety margins estimated from thermal-hydraulic code calculations are captured quantitatively by resorting to the order statistics and the bootstrap technique. The proposed framework of analysis is applied to the estimation of the safety margin, with its confidence interval, of the maximum fuel cladding temperature reached during a complete group distribution blockage scenario in a RBMK-1500 nuclear reactor.

Return on experience. Building tomorrow's safety rules

International Nuclear Information System (INIS)

Anon.

2013-01-01

Based on some experiments and witnesses, this set of articles outlines the importance of the return on experience in the evolution of process for improving the safety of nuclear installations. As an example, an article notably addresses the failures of Diesel engines which are used as spare generators in nuclear power stations: failures have been analysed and attributed to bearings, and more specifically to the degradation of their surface, and a third generation of bearings has been designed and produced. An article outlines that different industrials belonging to the nuclear sector as well as to other sectors share their returns on experience

Predicted thermal and stress environments in the vicinity of repository openings

International Nuclear Information System (INIS)

Bauer, S.J.; Hardy, M.P.; Lin, M.

1991-01-01

An understanding of the thermal and stress environment in the vicinity of repository openings is important for preclosure performance considerations and worker health and safety considerations for the proposed high-level radioactive waste repository at Yucca Mountain. This paper presents the results of two and three dimensional numerical analyses which have determined the thermal and stress environments for typical repository openings. In general, it is predicted that openings close to heat sources attain high temperatures and experience a significant stress increase. Openings away from heat sources experience more uniform temperature changes and experience a stress change which results in part from a far-field thermal loading

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

International Nuclear Information System (INIS)

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

1978-02-01

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

Thermal control surfaces on the MSFC LDEF experiments

International Nuclear Information System (INIS)

Wilkes, D.R.; Whitaker, A.F.; Zwiener, J.M.; Linton, R.C.; Shular, D.; Peters, P.N.; Gregory, J.C.

1992-01-01

There were five Marshall Space Flight Center (MSFC) experiments on the LDEF. Each of those experiments carried thermal control surfaces either as test samples or as operational surfaces. These materials experienced varying degrees of mechanical and optical damage. Some materials were virtually unchanged by the extended exposure while others suffered extensive degradation. The synergistic effects due to the constituents of the space environment are evident in the diversity of these material changes. The sample complement for the MSFC experiments is described along with results of the continuing analyses efforts

Safety climate mapping in a nuclear power plant - an experience sharing

International Nuclear Information System (INIS)

Vincy, M.U.; Varshney, Aloke; Khot, Pankaj

2016-01-01

In this paper the author discusses the experience gained in safety climate measurement of an Indian nuclear power plant. Safety performance is increasingly part of an organisation's sustainable development. Nuclear power stations are falling under the category 'high reliability' industries in the world as far as work safety is concerned. Both the research and the practical experience continually point to two underlying factors that drive safety outcomes: the quality of an organisation's leadership and the resulting culture. After years of development in safety technology and safety management system in the industry, management of nuclear industry world over has come to recognize that safety culture has to be addressed if high standards of health and safety are to be maintained. Therefore, nuclear industries in India have been carrying out measurement of safety climate for more than ten years. The objectives of the study are to examine people's values, attitude, perception, competencies, and patterns of behaviour that determine the commitment to, and effectiveness of health and safety management in the industry based on questionnaires survey and their analysis. A questionnaire, consists of 66 statements with 11 attributes, was designed to seek the views of managers, supervisors and front line workers on key aspects of the safety culture. Each of the discrete group was also classified according to their role in the organisation

Experiment study on thermal mixing performance of HTR-PM reactor outlet

Energy Technology Data Exchange (ETDEWEB)

Zhou, Yangping, E-mail: zhouyp@mail.tsinghua.edu.cn [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, the Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of Education, Tsinghua University, Beijing 100084 (China); Hao, Pengfei [School of Aerospace, Tsinghua University, Beijing 100084 (China); Li, Fu; Shi, Lei [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, the Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of Education, Tsinghua University, Beijing 100084 (China); He, Feng [School of Aerospace, Tsinghua University, Beijing 100084 (China); Dong, Yujie; Zhang, Zuoyi [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, the Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of Education, Tsinghua University, Beijing 100084 (China)

2016-09-15

A model experiment is proposed to investigate the thermal mixing performance of HTR-PM reactor outlet. The design of the test facility is introduced, which is set at a scale of 1:2.5 comparing with the design of thermal mixing structure at HTR-PM reactor outlet. The test facility using air as its flow media includes inlet pipe system, electric heaters, main mixing structure, hot gas duct, exhaust pipe system and I&C system. Experiments are conducted on the test facility and the values of thermal-fluid parameters are collected and analyzed, which include the temperature, pressure and velocity of the flow as well as the temperature of the tube wall. The analysis results show the mixing efficiency of the test facility is higher than that required by the steam generator of HTR-PM, which indicates that the thermal mixing structure of HTR-PM fulfills its design requirement.

Experience gained in enhancing operational safety at ComEd's nuclear power plants

International Nuclear Information System (INIS)

Elias, D.

1997-01-01

The following aspects of experience gained in enhancing operational safety at Comed's nuclear power plants are discussed: nuclear safety policy; centralization/decentralization; typical nuclear operating organization; safety review boards; human performance enhancement; elements of effective nuclear oversight

Benchmarking criticality safety calculations with subcritical experiments

International Nuclear Information System (INIS)

Mihalczo, J.T.

1984-06-01

Calculation of the neutron multiplication factor at delayed criticality may be necessary for benchmarking calculations but it may not be sufficient. The use of subcritical experiments to benchmark criticality safety calculations could result in substantial savings in fuel material costs for experiments. In some cases subcritical configurations could be used to benchmark calculations where sufficient fuel to achieve delayed criticality is not available. By performing a variety of measurements with subcritical configurations, much detailed information can be obtained which can be compared directly with calculations. This paper discusses several measurements that can be performed with subcritical assemblies and presents examples that include comparisons between calculation and experiment where possible. Where not, examples from critical experiments have been used but the measurement methods could also be used for subcritical experiments

Trend analysis of troubles caused by thermal-hydraulic phenomena at nuclear power plants

International Nuclear Information System (INIS)

Komatsu, Teruo

2010-01-01

The Institute of Nuclear Safety System (INSS) is promoting researches to improve the safety and reliability of nuclear power plants. In the present study, our attention was focused on troubles attributed to thermal-hydraulic phenomena in particular, trend analysis were carried out to learn lessons from these troubles and to prevent their recurrence. Through our survey, we found the following two points. First, many thermal-hydraulics related troubles can be attributed to design faults, since we found some events in foreign countries took place after inadequate facility renovation. To ensure appropriate design verification, it is important to take account of state-of-the-art science and technology and at the same time to pay attention to the compatibility with the initial design concept. Second point, thermal-hydraulic related troubles are common and recurrent to nuclear power plants worldwide. Japanese utilities are planning to introduce some of overseas experiences to their plants, such as power uprate and renovations of aged facilities. It is important to learn lessons from experiences paying close attention continuously to overseas trouble events, including thermal-hydraulics related events, and to use them to improve safety and reliability of nuclear power plants. (author)

Thermal stress-dependent dilation of concrete

International Nuclear Information System (INIS)

Pfeiffer, P.A.; Marchertas, A.H.

1984-01-01

Recent studies in nuclear fast reactor safety consider the possibility of concrete containment being subjected to extremely severe environmental conditions. Certain safety scenarios subject the concrete to very high temperatures hence raising the concern of containment integrity. Some of the main detrimental effects of high temperature on concrete are: reduction of strength, redistribution of moisture and etc. Consequently, analytical prediction of concrete response under the high temperature conditions becomes very complex. A rather simple but important experiment of concrete at high temperatures was conducted by Anderberg and Thelandersson. The test samples were small so that moisture was free to evaporate with no appreciable gradient as the temperature increased. Their results revealed that good correlation with analysis could be obtained if thermal expansion was made a function of both temperature and stress. The method of relating the thermal strain to temperature and stress has been integrated into the TEMP-STRESS code. Thus, high temperature concrete computational capability is now available for thermal-stress calculations

Operating experience feedback from safety significant events at research reactors

Energy Technology Data Exchange (ETDEWEB)

Shokr, A.M. [Atomic Energy Authority, Abouzabal (Egypt). Egypt Second Research Reactor; Rao, D. [Bhabha Atomic Research Centre, Mumbai (India)

2015-05-15

Operating experience feedback is an effective mechanism to provide lessons learned from the events and the associated corrective actions to prevent recurrence of events, resulting in improving safety in the nuclear installations. This paper analyzes the events of safety significance that have been occurred at research reactors and discusses the root causes and lessons learned from these events. Insights from literature on events at research reactors and feedback from events at nuclear power plants that are relevant to research reactors are also presented along with discussions. The results of the analysis showed the importance of communication of safety information and exchange of operating experience are vital to prevent reoccurrences of events. The analysis showed also the need for continued attention to human factors and training of operating personnel, and the need for establishing systematic ageing management programmes of reactor facilities, and programmes for safety management of handling of nuclear fuel, core components, and experimental devices.

Operating experience feedback from safety significant events at research reactors

International Nuclear Information System (INIS)

Shokr, A.M.

2015-01-01

Operating experience feedback is an effective mechanism to provide lessons learned from the events and the associated corrective actions to prevent recurrence of events, resulting in improving safety in the nuclear installations. This paper analyzes the events of safety significance that have been occurred at research reactors and discusses the root causes and lessons learned from these events. Insights from literature on events at research reactors and feedback from events at nuclear power plants that are relevant to research reactors are also presented along with discussions. The results of the analysis showed the importance of communication of safety information and exchange of operating experience are vital to prevent reoccurrences of events. The analysis showed also the need for continued attention to human factors and training of operating personnel, and the need for establishing systematic ageing management programmes of reactor facilities, and programmes for safety management of handling of nuclear fuel, core components, and experimental devices.

Safety of nuclear power plants: Operation. Safety requirements

International Nuclear Information System (INIS)

2004-01-01

The safety of a nuclear power plant is ensured by means of its proper siting, design, construction and commissioning, followed by the proper management and operation of the plant. In a later phase, proper decommissioning is required. This Safety Requirements publication supersedes the Code on the Safety of Nuclear Power Plants: Operation, which was issued in 1988 as Safety Series No. 50-C-O (Rev. 1). The purpose of this revision was: to restructure Safety Series No. 50-C-O (Rev. 1) in the light of the basic objectives, concepts and principles in the Safety Fundamentals publication The Safety of Nuclear Installations. To be consistent with the requirements of the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources. And to reflect current practice and new concepts and technical developments. Guidance on fulfillment of these Safety Requirements may be found in the appropriate Safety Guides relating to plant operation. The objective of this publication is to establish the requirements which, in the light of experience and the present state of technology, must be satisfied to ensure the safe operation of nuclear power plants. These requirements are governed by the basic objectives, concepts and principles that are presented in the Safety Fundamentals publication The Safety of Nuclear Installations. This publication deals with matters specific to the safe operation of land based stationary thermal neutron nuclear power plants, and also covers their commissioning and subsequent decommissioning

Safety of nuclear power plants: Operation. Safety requirements

International Nuclear Information System (INIS)

2003-01-01

The safety of a nuclear power plant is ensured by means of its proper siting, design, construction and commissioning, followed by the proper management and operation of the plant. In a later phase, proper decommissioning is required. This Safety Requirements publication supersedes the Code on the Safety of Nuclear Power Plants: Operation, which was issued in 1988 as Safety Series No. 50-C-O (Rev. 1). The purpose of this revision was: to restructure Safety Series No. 50-C-O (Rev. 1) in the light of the basic objectives, concepts and principles in the Safety Fundamentals publication The Safety of Nuclear Installations. To be consistent with the requirements of the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources. And to reflect current practice and new concepts and technical developments. Guidance on fulfillment of these Safety Requirements may be found in the appropriate Safety Guides relating to plant operation. The objective of this publication is to establish the requirements which, in the light of experience and the present state of technology, must be satisfied to ensure the safe operation of nuclear power plants. These requirements are governed by the basic objectives, concepts and principles that are presented in the Safety Fundamentals publication The Safety of Nuclear Installations. This publication deals with matters specific to the safe operation of land based stationary thermal neutron nuclear power plants, and also covers their commissioning and subsequent decommissioning

Safety of nuclear power plants: Operation. Safety requirements

International Nuclear Information System (INIS)

2000-01-01

The safety of a nuclear power plant is ensured by means of its proper siting, design, construction and commissioning, followed by the proper management and operation of the plant. In a later phase, proper decommissioning is required. This Safety Requirements publication supersedes the Code on the Safety of Nuclear Power Plants: Operation, which was issued in 1988 as Safety Series No. 50-C-O (Rev. 1). The purpose of this revision was: to restructure Safety Series No. 50-C-O (Rev. 1) in the light of the basic objectives, concepts and principles in the Safety Fundamentals publication The Safety of Nuclear Installations; to be consistent with the requirements of the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources; and to reflect current practice and new concepts and technical developments. Guidance on fulfillment of these Safety Requirements may be found in the appropriate Safety Guides relating to plant operation. The objective of this publication is to establish the requirements which, in the light of experience and the present state of technology, must be satisfied to ensure the safe operation of nuclear power plants. These requirements are governed by the basic objectives, concepts and principles that are presented in the Safety Fundamentals publication The Safety of Nuclear Installations. This publication deals with matters specific to the safe operation of land based stationary thermal neutron nuclear power plants, and also covers their commissioning and subsequent decommissioning

"Nursing Students Assaulted": Considering Student Safety in Community-Focused Experiences.

Science.gov (United States)

Maneval, Rhonda E; Kurz, Jane

2016-01-01

Community nursing experiences for undergraduate students have progressed beyond community-based home visits to a wide array of community-focused experiences in neighborhood-based centers, clinics, shelters, and schools. Our Bachelor of Science in Nursing program chose to use sites situated within neighborhoods close to campus in order to promote student and faculty engagement in the local community. These neighborhood sites provide opportunities for students to deliver nursing services to underserved and vulnerable populations experiencing poverty and health disparities. Some of these neighborhoods are designated as high crime areas that may potentially increase the risk of harm to students and faculty. There is a need to acknowledge the risk to personal safety and to proactively create policies and guidelines to reduce potential harm to students engaged in community-focused experiences. When a group of baccalaureate nursing students was assaulted while walking to a neighborhood clinic, the faculty was challenged as how to respond given the lack of policies and guidelines. Through our experience, we share strategies to promote personal safety for students and recommend transparency by administrators regarding potential safety risks to students engaged in community-focused fieldwork activities. Copyright © 2016 Elsevier Inc. All rights reserved.

Experience base for Radioactive Waste Thermal Processing Systems: A preliminary survey

International Nuclear Information System (INIS)

Mayberry, J.; Geimer, R.; Gillins, R.; Steverson, E.M.; Dalton, D.; Anderson, G.L.

1992-04-01

In the process of considering thermal technologies for potential treatment of the Idaho National Engineering Laboratory mixed transuranic contaminated wastes, a preliminary survey of the experience base available from Radioactive Waste Thermal Processing Systems is reported. A list of known commercial radioactive waste facilities in the United States and some international thermal treatment facilities are provided. Survey focus is upon the US Department of Energy thermal treatment facilities. A brief facility description and a preliminary summary of facility status, and problems experienced is provided for a selected subset of the DOE facilities

Comments on thermal runaway experiments in sub-ignition tokamaks

International Nuclear Information System (INIS)

Yamazaki, K.

1982-09-01

Justification of deuterium-tritium operations is discussed from the physics viewpoint and optimal thermal runaway experiments in high-field, high-density compact tokamaks are suggested within the minimization of the induced radioactivation. (author)

The Importance of a Thermal Manikin as Source and Obstacle in Full-Scale Experiments

DEFF Research Database (Denmark)

Nielsen, Peter V.

The thermal manikin is normally introduced at indoor environmental measurements to obtain detailed information on thermal comfort and air quality around a person. This paper deals with the opposite situation where manikins are introduced as sources and obstacles in order to obtain reasonable...... boundary conditions in experiments with the indoor environment. In other words, how will people influence the surroundings instead of how will the surroundings influence people? The use of thermal manikins in an experiment will of course take both situations into account, however, in some experiments...

Implications of passive safety based on historical industrial experience

International Nuclear Information System (INIS)

Forsberg, C.W.

1988-01-01

In the past decade, there have been multiple proposals for applying different technologies to achieve passively safe light water reactors (LWRs). A key question for all such concepts is, ''What are the gains in safety, costs, and reliability for passive safety systems.'' Using several types of historical data, estimates have been made of gains from passive safety and operating systems, which are independent of technology. Proposals for passive safety in reactors usually have three characteristics: (1) Passive systems with no moving mechanical parts, (2) systems with far fewer components and (3) more stringent design criteria for safety-related and process systems. Each characteristic reduces the potential for an accident and may increase plant reliability. This paper addresses gains from items (1) and (2). Passive systems often allow adoption of more rigorous design criteria which would be either impossible or economically unfeasible for active systems. This important characteristic of passive safety systems cannot be easily addressed using historical industrial experience

The importance for nuclear safety of efficient feedback of operational experience

International Nuclear Information System (INIS)

1987-09-01

Experience of practical operation is a valuable source of information for improving and optimizing the safety and reliability of nuclear power plants. Therefore it is essential to collect information on abnormal events occurring at plants during operation and on all deviations from normal performance by systems and personnel that could be precursors of accidents. For this purpose it is necessary to establish hierarchical systems to feedback operational safety experience at utility, national and international levels and to make these systems as effective as possible. The present report attempts to identify the safety objectives of these systems, to analyse the difficulties presently encountered and to suggest possible improvements

Evaluation of operating experience for early recognition of deteriorating safety performance

International Nuclear Information System (INIS)

Beckmerhagen, I.A.; Berg, H.P.

2004-01-01

One of the most difficult challenges facing nuclear power plants is to recognize the early signs of degrading safety performance before regulatory requirements are imposed or serious incidents or accidents occur. Today, the nuclear industry is striving for collecting more information on occurrences that could improve the operational safety performance. To achieve this, the reporting threshold has been lowered from incidents to anomalies with minor or no impact to safety. Industry experience (also outside nuclear industry) has shown that these are typical issues which should be considered when looking for such early warning signs. Therefore, it is important that nuclear power plant operators have the capability to trend, analyse and recognize early warning signs of deteriorating performance. It is necessary that plant operators are sensitive to these warning signs which may not be immediately evident. Reviewing operating experience is one of the main tasks for plant operators in their daily activities. Therefore, self assessment should be at the centre of any operational safety performance programme. One way of applying a self assessment program is through the following four basic elements: operational data, events, safety basis, and related experience. This approach will be described in the paper in more details. (authors)

Forschungszentrum Rossendorf, Institute of Safety Research. Annual report 2003

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-07-01

The work of the institute is directed to the assessment and enhancement of the safety of technical plants and to the increase of the efficiency and environmental sustainability of those facilities. Subjects of investigation are equally nuclear plants and installations of process industries. To achieve the goals mentioned, the institute is mainly engaged in the scientific fields of thermal fluid dynamics including magneto-hydrodynamics (MHD) and materials sciences. In 2003, the ISR worked on the following main scientific projects. Sub-programme: Plant and Rector Safety. Project: accident analysis of nuclear reactors, safety of materials and components, particle and radiation transport, safety and efficiency of chemical processes. Sub-programme: Thermal Fluid Dynamics. Project: magneto-hydrodynamics, thermal fluid dynamics of multiphase systems. Considerable progress could also be achieved in the CFD simulation of two-phase flows. New approaches for the forces acting on steam bubbles in a water flow could be developed and implemented into the CFX code in close cooperation with the CFX developer ANSYS/CFX. The qualified models allow to simulate the evolution of bubble size specific radial void distribution profiles along the flow path. These theoretical studies and the related experiments at the Rossendorf TOPFLOW test facility represent an important part of the German CFD network that aims at the improvement of thermal hydraulic calculation methods in reactor safety. (orig.)

Forschungszentrum Rossendorf, Institute of Safety Research. Annual report 2003

International Nuclear Information System (INIS)

2004-01-01

The work of the institute is directed to the assessment and enhancement of the safety of technical plants and to the increase of the efficiency and environmental sustainability of those facilities. Subjects of investigation are equally nuclear plants and installations of process industries. To achieve the goals mentioned, the institute is mainly engaged in the scientific fields of thermal fluid dynamics including magneto-hydrodynamics (MHD) and materials sciences. In 2003, the ISR worked on the following main scientific projects. Sub-programme: Plant and Rector Safety. Project: accident analysis of nuclear reactors, safety of materials and components, particle and radiation transport, safety and efficiency of chemical processes. Sub-programme: Thermal Fluid Dynamics. Project: magneto-hydrodynamics, thermal fluid dynamics of multiphase systems. Considerable progress could also be achieved in the CFD simulation of two-phase flows. New approaches for the forces acting on steam bubbles in a water flow could be developed and implemented into the CFX code in close cooperation with the CFX developer ANSYS/CFX. The qualified models allow to simulate the evolution of bubble size specific radial void distribution profiles along the flow path. These theoretical studies and the related experiments at the Rossendorf TOPFLOW test facility represent an important part of the German CFD network that aims at the improvement of thermal hydraulic calculation methods in reactor safety. (orig.)

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

Status and topics of thermal-hydraulic analysis for next-generation LWRs with passive safety systems

International Nuclear Information System (INIS)

Aritomi, Masanori; Ohnuki, Akira; Arai, Kenji; Kikuta, Michitaka; Yonomoto, Taisuke; Araya, Fumimasa; Akimoto, Hajime

1999-01-01

For increasing of electric power demand and reducing of carbon dioxide exhaust in the 21st century, studies of the next-generation light water reactor (LWR) with passive safety systems are developing in the world: AP-600 (by Westing House Co.); SBWR (by General Electric Co.); SWR1000 (by Siemens Co.); NP21 (by Mitsubishi Heavy Industry Co., et al.); JPSR (by JAERI). The passive equipment using natural circulation and natural convection are installed in the passive safety system, instead of active safety equipment, such as pumps, etc. It remains still as a important issue, however, to verify the reliability on the functions of the passive equipment, since that the driving forces of the passive equipment are small at comparison with the active safety equipment. The various subjects of thermal-hydraulic analysis for the next-generation light water reactors, such as temperature stratification in the passive safety systems, vapor condensation in the mixture of non-condensable gases and the interactions of the passive safety system with the primary cooling system, are illustrated and discussed in the paper. (M. Suetake)

Introduction to Safety Analysis Approach for Research Reactors

International Nuclear Information System (INIS)

Park, Suki

2016-01-01

The research reactors have a wide variety in terms of thermal powers, coolants, moderators, reflectors, fuels, reactor tanks and pools, flow direction in the core, and the operating pressure and temperature of the cooling system. Around 110 research reactors have a thermal power greater than 1 MW. This paper introduces a general approach to safety analysis for research reactors and deals with the experience of safety analysis on a 10 MW research reactor with an open-pool and open-tank reactor and a downward flow in the reactor core during normal operation. The general approach to safety analysis for research reactors is described and the design features of a typical open-pool and open-tank type reactor are discussed. The representative events expected in research reactors are investigated. The reactor responses and the thermal hydraulic behavior to the events are presented and discussed. From the minimum CHFR and the maximum fuel temperature calculated, it is ensured that the fuel is not damaged in the step insertion of reactivity by 1.8 mk and the failure of all primary pumps for the reactor with a 10 MW thermal power and downward core flow

Experiments on thermal conductivity in buffer materials for geologic repository

International Nuclear Information System (INIS)

Kanno, T.; Yano, T.; Wakamatsu, H.; Matsushima, E.

1989-01-01

Engineered barriers for geologic disposal for HLW are planned to consist of canister, overpack and buffer elements. One of important physical characteristics of buffer materials is determining temperature profiles within the near field in a repository. Buffer materials require high thermal conductivity to disperse radiogenic heat away to the host rock. As the buffer materials, compacted blocks of the mixture of sodium bentonite and sand have been the most promising candidate in some countries, e.g. Sweden, Switzerland and Japan. The authors have been carrying out a series of thermal dispersion experiments to evaluate thermal conductivity of bentonite/quartz sand blocks. In this study, the following two factors considered to affect thermal properties of the near field were examined: effective thermal conductivities of buffer materials, and heat transfer characteristics of the gap between overpack and buffer materials

Primary system thermal hydraulics of future Indian fast reactors

Energy Technology Data Exchange (ETDEWEB)

Velusamy, K., E-mail: kvelu@igcar.gov.in [Thermal Hydraulics Section, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Natesan, K.; Maity, Ram Kumar; Asokkumar, M.; Baskar, R. Arul; Rajendrakumar, M.; Sarathy, U. Partha; Selvaraj, P.; Chellapandi, P. [Thermal Hydraulics Section, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Kumar, G. Senthil; Jebaraj, C. [AU-FRG Centre for CAD/CAM, Anna University, Chennai 600 025 (India)

2015-12-01

Highlights: • We present innovative design options proposed for future Indian fast reactor. • These options have been validated by extensive CFD simulations. • Hotspot factors in fuel subassembly are predicted by parallel CFD simulations. • Significant safety improvement in the thermal hydraulic design is quantified. - Abstract: As a follow-up to PFBR (Indian prototype fast breeder reactor), many FBRs of 500 MWe capacity are planned. The focus of these future FBRs is improved economy and enhanced safety. They are envisaged to have a twin-unit concept. Design and construction experiences gained from PFBR project have provided motivation to achieve an optimized design for future FBRs with significant design changes for many critical components. Some of the design changes include, (i) provision of four primary pipes per primary sodium pump, (ii) inner vessel with single torus lower part, (iii) dome shape roof slab supported on reactor vault, (iv) machined thick plate rotating plugs, (v) reduced main vessel diameter with narrow-gap cooling baffles and (vi) safety vessel integrated with reactor vault. This paper covers thermal hydraulic design validation of the chosen options with respect to hot and cold pool thermal hydraulics, flow requirement for main vessel cooling, inner vessel temperature distribution, safety analysis of primary pipe rupture event, adequacy of decay heat removal capacity by natural convection cooling, cold pool transient thermal loads and thermal management of top shield and reactor vault.

Large-scale thermal-shock experiments with clad and unclad steel cylinders

International Nuclear Information System (INIS)

Cheverton, R.D.

1992-01-01

Flaw behavior trends associated with pressurized-thermal-shock (PTS) loading of pressurized-water-reactor pressure vessels have been under investigation at the Oak Ridge National Laboratory for nearly 20 years. During that time, twelve thermal-shock experiments with thick-walled (152 mm) steel cylinders were conducted as a part of the investigations. The first eight experiments were conducted with unclad cylinders initially containing shallow (8--19 mm) two-dimensional and semicircular inner-surface flaws. These experiments demonstrated, in good agreement with linear elastic fracture mechanics, crack initiation and arrest, a series of initiation/arrest events with deep penetration of the wall, long crack jumps, arrest with the stress intensity factor (K I ) increasing with crack depth, extensive surface extension of an initially short and shallow (semicircular) flaw, and warm prestressing with K I ≤ 0. The remaining four experiments were conducted with clad cylinders containing initially shallow (19--24 mm) semielliptical subclad and surface flaws at the inner surface. In the first of these experiments one of six equally spaced (60 degrees) open-quotes identicalclose quotes subclad flaws extended nearly the length of the cylinder (1,220 mm) beneath the cladding (no crack extension into the cladding) and nearly 50% of the wall, radially. For the final experiment, four of the semielliptical subclad flaws that had not propagated previously were converted to surface flaws, and they experienced extensive extension beneath the cladding with no cracking of the cladding. Information from this series of thermal-shock experiments is being used in the evaluation of the PTS issue

Workshop on gate valve pressure locking and thermal binding

International Nuclear Information System (INIS)

Brown, E.J.

1995-07-01

The purpose of the Workshop on Gate Valve Pressure Locking and Thermal Binding was to discuss pressure locking and thermal binding issues that could lead to inoperable gate valves in both boiling water and pressurized water reactors. The goal was to foster exchange of information to develop the technical bases to understand the phenomena, identify the components that are susceptible, discuss actual events, discuss the safety significance, and illustrate known corrective actions that can prevent or limit the occurrence of pressure locking or thermal binding. The presentations were structured to cover U.S. Nuclear Regulatory Commission staff evaluation of operating experience and planned regulatory activity; industry discussions of specific events, including foreign experience, and efforts to determine causes and alleviate the affects; and valve vendor experience and recommended corrective action. The discussions indicated that identifying valves susceptible to pressure locking and thermal binding was a complex process involving knowledge of components, systems, and plant operations. The corrective action options are varied and straightforward

Nurses' experiences and perspectives on medication safety practices: an explorative qualitative study.

Science.gov (United States)

Smeulers, Marian; Onderwater, Astrid T; van Zwieten, Myra C B; Vermeulen, Hester

2014-04-01

To explore nurses' experiences with and perspectives on preventing medication administration errors. Insight into nurses' experiences with and perspectives on preventing medication administration errors is important and can be utilised to tailor and implement safety practices. A qualitative interview study of 20 nurses in an academic medical centre was conducted between March and December of 2011. Three themes emerged from this study: (1) nurses' roles and responsibilities in medication safety: aside from safe preparation and administration, the clinical reasoning of nurses is essential for medication safety; (2) nurses' ability to work safely: knowledge of risks and nurses' work circumstances influence their ability to work safely; and (3) nurses' acceptance of safety practices: advantages, feasibility and appropriateness are important incentives for acceptance of a safety practice. Nurses' experiences coincide with the assumption that they are in a pre-eminent position to enable safe medication management; however, their ability to adequately perform this role depends on sufficient knowledge to assess the risks of medication administration and on the circumstances in which they work. Safe medication management requires a learning climate and professional practice environment that enables further development of professional nursing skills and knowledge. © 2014 John Wiley & Sons Ltd.

Safety issues for superconducting fusion magnets

International Nuclear Information System (INIS)

Hsieh, S.Y.; Reich, M.; Powell, J.R.

1978-01-01

Safety issues for future superconducting fusion magnet systems are examined. It is found that safety and failure experience with existing superconducting magnets is not very applicable to predictions as to the safety and reliability of fusion magnets. Such predictions will have to depend on analysis and judgement for many years to come, rather than on accumulated experience. A number of generic potential structural, thermal-hydraulic, and electrical safety problems are identified and analyzed. Prevention of quenches and non-uniform temperature distributions, if quenches should occur, is of great importance, since such events can trigger processes which lead to magnet damage or failure. Engineered safety features will be necessary for fusion magnets. Two of these, an energy dispersion system and external coil containment, appear capable of reducing the probability of coil disruption to very low levels. However, they do not prevent loss of function accidents which are of economic concern. Elaborate detector, temperature equalization, and energy removal systems will be required to minimize the chances of loss of function accidents

A new approach to designing reduced scale thermal-hydraulic experiments

International Nuclear Information System (INIS)

Lapa, Celso M.F.; Sampaio, Paulo A.B. de; Pereira, Claudio M.N.A.

2004-01-01

Reduced scale experiments are often employed in engineering because they are much cheaper than real scale testing. Unfortunately, though, it is difficult to design a thermal-hydraulic circuit or equipment in reduced scale capable of reproducing, both accurately and simultaneously, all the physical phenomena that occur in real scale and operating conditions. This paper presents a methodology to designing thermal-hydraulic experiments in reduced scale based on setting up a constrained optimization problem that is solved using genetic algorithms (GAs). In order to demonstrate the application of the methodology proposed, we performed some investigations in the design of a heater aimed to simulate the transport of heat and momentum in the core of a pressurized water reactor (PWR) at 100% of nominal power and non-accident operating conditions. The results obtained show that the proposed methodology is a promising approach for designing reduced scale experiments

COMMIX analysis of four constant flow thermal upramp experiments performed in a thermal hydraulic model of an advanced LMR

International Nuclear Information System (INIS)

Yarlagadda, B.S.

1989-04-01

The three-dimensional thermal hydraulics computer code COMMIX-1AR was used to analyze four constant flow thermal upramp experiments performed in the thermal hydraulic model of an advanced LMR. An objective of these analyses was the validation of COMMIX-1AR for buoyancy affected flows. The COMMIX calculated temperature histories of some thermocouples in the model were compared with the corresponding measured data. The conclusions of this work are presented. 3 refs., 5 figs

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

Experimental study on the thermal stratification in the branch of NPP

Energy Technology Data Exchange (ETDEWEB)

Kim, Sang Nyung; Hwang, Seong Hong [Kyunghee Univ., Seoul (Korea, Republic of)

2004-02-15

As more experience is accumulated in the operation of existing nuclear power plants, the long term effects of thermal-hydraulic phenomena, unaccounted in the original designs, have been observed. One such phenomenon is thermal stratification, which has caused through-wall cracks, thermal fatigue, unexpected piping displacements and pipe support damage. Thermal stratification is a phenomenon as temperature layers are formed in the component or piping due to the density difference between hot and cold water. The thermal stratification phenomena in nuclear power plant observed in the pressurizer surge line, and in the piping of feedwater system, Safety Injection System(SIS), residual heat removal system (or shutdown cooling system), and chemical and volume control system during the design transients. A set of experiment has been performed to predict the temperature distribution in the branch piping of nuclear power plant(Ulchin unit 3 and 4) due to the turbulent penetration, the heat transfer through valve disk and valve leakage. The test facility scaled down to 1/10 has been designed and constructed to simulate the thermal stratification in the piping of safety injection system and shutdown cooling system of Ulchin 3 and 4. The experimental results show that the turbulent penetration depth could be : extended to the end of the vertical pipe, and thermal stratification due to the heat transfer through the valve disk to the end of horizontal pipe behind the valve disk. Finally, thermal stratification could effected by the location of valve leakage.

Use of operational experience in fire safety assessment of nuclear power plants

International Nuclear Information System (INIS)

2000-01-01

collection of data related to fire safety occurrences in NPPs, the so called operational experience and the use of such operational experience in NPPs. This report provides good practice information on data needs, data reporting requirements and some advice on database features. In addition, this publication provides information on the applications of fire related operational experience, highlighting their benefits. This publication has been developed to complement other IAEA publications related to fire safety analysis within the framework of the IAEA programme of fire safety

Experience gained in enhancing operational safety at ComEd`s nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Elias, D [Commonwealth Edison Co. (United States)

1997-09-01

The following aspects of experience gained in enhancing operational safety at Comed`s nuclear power plants are discussed: nuclear safety policy; centralization/decentralization; typical nuclear operating organization; safety review boards; human performance enhancement; elements of effective nuclear oversight.

PReSaFe: A model of barriers and facilitators to patients providing feedback on experiences of safety.

Science.gov (United States)

De Brún, Aoife; Heavey, Emily; Waring, Justin; Dawson, Pamela; Scott, Jason

2017-08-01

The importance of involving patients in reporting on safety is increasingly recognized. Whilst studies have identified barriers to clinician incident reporting, few have explored barriers and facilitators to patient reporting of safety experiences. This paper explores patient perspectives on providing feedback on safety experiences. Patients (n=28) were invited to take part in semi-structured interviews when given a survey about their experiences of safety following hospital discharge. Transcripts were thematically analysed using NVivo10. Patients were recruited from four hospitals in the UK. Three themes were identified as barriers and facilitators to patient involvement in providing feedback on their safety experiences. The first, cognitive-cultural, found that whilst safety was a priority for most, some felt the term was not relevant to them because safety was the "default" position, and/or because safety could not be disentangled from the overall experience of care. The structural-procedural theme indicated that reporting was facilitated when patients saw the process as straightforward, but that disinclination or perceived inability to provide feedback was a barrier. Finally, learning and change illustrated that perception of the impact of feedback could facilitate or inhibit reporting. When collecting patient feedback on experiences of safety, it is important to consider what may help or hinder this process, beyond the process alone. We present a staged model of prerequisite barriers and facilitators and hypothesize that each stage needs to be achieved for patients to provide feedback on safety experiences. Implications for collecting meaningful data on patients' safety experiences are considered. © 2016 The Authors. Health Expectations Published by John Wiley & Sons Ltd.

Presurized water reactor safety approach and analysis. From conception to experience feedback

International Nuclear Information System (INIS)

Libmann, J.

1987-04-01

This report deals in ten chapters, with the following subjects: 1. Safety approach methods; 2. Study of accidents; 3. Safety analysis; 4. Study of internal aggressions or those involved by the site; 5. Consideration of complementary situations; 6. Three Mile Island accident; 7. Safety during operation and experience feedback; 8. An example of analysis: steam generator closure plug; 9. Probabilistic safety evaluation; 10. Chernobyl accident. 30 refs [fr

Thermal Runaways in LHC Interconnections: Experiments

CERN Document Server

Willering, G P; Bottura, L; Scheuerlein, C; Verweij, A P

2011-01-01

The incident in the LHC in September 2008 occurred in an interconnection between two magnets of the 13 kA dipole circuit. This event was traced to a defect in one of the soldered joints between two superconducting cables stabilized by a copper busbar. Further investigation revealed defective joints of other types. A combination of (1) a poor contact between the superconducting cable and the copper stabilizer and (2) an electrical discontinuity in the stabilizer at the level of the connection can lead to an unprotected quench of the busbar. Once the heating power in the unprotected superconducting cable exceeds the heat removal capacity a thermal run-away occurs, resulting in a fast melt-down of the non-stabilized cable. We have performed a thorough investigation of the conditions upon which a thermal run-away in the defect can occur. To this aim, we have prepared heavily instrumented samples with well-defined and controlled defects. In this paper we describe the experiment, and the analysis of the data, and w...

Modelling of thermal shock experiments of carbon based materials in JUDITH

Science.gov (United States)

Ogorodnikova, O. V.; Pestchanyi, S.; Koza, Y.; Linke, J.

2005-03-01

The interaction of hot plasma with material in fusion devices can result in material erosion and irreversible damage. Carbon based materials are proposed for ITER divertor armour. To simulate carbon erosion under high heat fluxes, electron beam heating in the JUDITH facility has been used. In this paper, carbon erosion under energetic electron impact is modeled by the 3D thermomechanics code 'PEGASUS-3D'. The code is based on a crack generation induced by thermal stress. The particle emission observed in thermal shock experiments is a result of breaking bonds between grains caused by thermal stress. The comparison of calculations with experimental data from JUDITH shows good agreement for various incident power densities and pulse durations. A realistic mean failure stress has been found. Pre-heating of test specimens results in earlier onset of brittle destruction and enhanced particle loss in agreement with experiments.

Modelling of thermal shock experiments of carbon based materials in JUDITH

International Nuclear Information System (INIS)

Ogorodnikova, O.V.; Pestchanyi, S.; Koza, Y.; Linke, J.

2005-01-01

The interaction of hot plasma with material in fusion devices can result in material erosion and irreversible damage. Carbon based materials are proposed for ITER divertor armour. To simulate carbon erosion under high heat fluxes, electron beam heating in the JUDITH facility has been used. In this paper, carbon erosion under energetic electron impact is modeled by the 3D thermomechanics code 'PEGASUS-3D'. The code is based on a crack generation induced by thermal stress. The particle emission observed in thermal shock experiments is a result of breaking bonds between grains caused by thermal stress. The comparison of calculations with experimental data from JUDITH shows good agreement for various incident power densities and pulse durations. A realistic mean failure stress has been found. Pre-heating of test specimens results in earlier onset of brittle destruction and enhanced particle loss in agreement with experiments

Modelling of thermal shock experiments of carbon based materials in JUDITH

Energy Technology Data Exchange (ETDEWEB)

Ogorodnikova, O.V. [Forschungszentrum Juelich, EURATOM-Association, IWV-2, 52425 Juelich (Germany)]. E-mail: o.ogorodnikova@fz-juelich.de; Pestchanyi, S. [Forschungszentrum Karlsruhe, EURATOM-Associaton, IHM, 76021 Karlsruhe (Germany); Koza, Y. [Forschungszentrum Juelich, EURATOM-Association, IWV-2, 52425 Juelich (Germany); Linke, J. [Forschungszentrum Juelich, EURATOM-Association, IWV-2, 52425 Juelich (Germany)

2005-03-01

The interaction of hot plasma with material in fusion devices can result in material erosion and irreversible damage. Carbon based materials are proposed for ITER divertor armour. To simulate carbon erosion under high heat fluxes, electron beam heating in the JUDITH facility has been used. In this paper, carbon erosion under energetic electron impact is modeled by the 3D thermomechanics code 'PEGASUS-3D'. The code is based on a crack generation induced by thermal stress. The particle emission observed in thermal shock experiments is a result of breaking bonds between grains caused by thermal stress. The comparison of calculations with experimental data from JUDITH shows good agreement for various incident power densities and pulse durations. A realistic mean failure stress has been found. Pre-heating of test specimens results in earlier onset of brittle destruction and enhanced particle loss in agreement with experiments.

Additive model for thermal comfort generated by matrix experiment using orthogonal array

Energy Technology Data Exchange (ETDEWEB)

Hwang, Reuy-Lung [Department of Occupational Safety and Health, China Medical University, 91 Huseh-shin Road, Taichung 404 (China); Lin, Tzu-Ping [Department of Leisure Planning, National Formosa University, 64 Wen-hua Road, Huwei, Yunlin 632 (China); Liang, Han-Hsi [Department of Architecture, National United University, No. 1, Lien Da, Kung-Ching Li, Miaoli 360 (China); Yang, Kuan-Hsiug; Yeh, Tsung-Chyn [Department of Mechanical and Electro-Mechanical Engineering, National Sun Yet-Sen University, No. 91, Lien-hai Road, Kaohsiung (China)

2009-08-15

In addition to ensuring the thermal comfort of occupants, monitoring and controlling indoor thermal environments can reduce the energy consumed by air conditioning systems. This study develops an additive model for predicting thermal comfort with rapid and simple arithmetic calculations. The advantage of the additive model is its comprehensibility to administrators of air conditioning systems, who are unfamiliar with the PMV-PPD model but want to adjust an indoor environment to save energy without generating complaints of discomfort from occupants. In order to generate the additive model, a laboratory chamber experiment based on matrix experiment using orthogonal array, was performed. By applying the analysis of variance on observed thermal sensation votes and percentage of dissatisfaction, the factor effects of environmental variables that account for the additive model were determined. Additionally, the applicability of the PMV-PPD model in hot and humid climates is discussed in this study, based on experimental results. (author)

Plasma diagnostic techniques in thermal-barrier tandem-mirror fusion experiments

International Nuclear Information System (INIS)

Silver, E.H.; Clauser, J.F.; Carter, M.R.; Failor, B.H.; Foote, J.H.; Hornady, R.S.; James, R.A.; Lasnier, C.J.; Perkins, D.E.

1986-01-01

We review two classes of plasma diagnostic techniques used in thermal-barrier tandem-mirror fusion experiments. The emphasis of the first class is to study mirror-trapped electrons at the thermal-barrier location. The focus of the second class is to measure the spatial and temporal behavior of the plasma space potential at various axial locations. The design and operation of the instruments in these two categories are discussed and data that are representative of their performance is presented

Planning the Unplanned Experiment: Assessing the Efficacy of Standards for Safety Critical Software

Science.gov (United States)

Graydon, Patrick J.; Holloway, C. Michael

2015-01-01

We need well-founded means of determining whether software is t for use in safety-critical applications. While software in industries such as aviation has an excellent safety record, the fact that software aws have contributed to deaths illustrates the need for justi ably high con dence in software. It is often argued that software is t for safety-critical use because it conforms to a standard for software in safety-critical systems. But little is known about whether such standards `work.' Reliance upon a standard without knowing whether it works is an experiment; without collecting data to assess the standard, this experiment is unplanned. This paper reports on a workshop intended to explore how standards could practicably be assessed. Planning the Unplanned Experiment: Assessing the Ecacy of Standards for Safety Critical Software (AESSCS) was held on 13 May 2014 in conjunction with the European Dependable Computing Conference (EDCC). We summarize and elaborate on the workshop's discussion of the topic, including both the presented positions and the dialogue that ensued.

Proceedings of 2nd PHWR operating safety experience meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

1991-04-01

Papers presented during the eight sessions of the meeting were devoted to the impact of PHWR operating experience on design of civil structures (reactor building integrity); operating experiences related to pressure tubes, nuclear steam supply system, plant stability; reactor maintenance and control systems, reactor operational safety. Some events concerned with reactor shutdown due to power failures are described, as well as action undertaken to prevent major damage.

Proceedings of 2nd PHWR operating safety experience meeting

International Nuclear Information System (INIS)

1991-04-01

Papers presented during the eight sessions of the meeting were devoted to the impact of PHWR operating experience on design of civil structures (reactor building integrity); operating experiences related to pressure tubes, nuclear steam supply system, plant stability; reactor maintenance and control systems, reactor operational safety. Some events concerned with reactor shutdown due to power failures are described, as well as action undertaken to prevent major damage

The SESAME project. State of the art liquid metal thermal hydraulics and beyond

Energy Technology Data Exchange (ETDEWEB)

Roelofs, F.; Shams, A.; Batta, A.; Moreau, V.; Di Piazza, I.; Gerschenfeld, A.; Planquart, P.; Tarantino, M. [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands)

2017-08-15

The European Sustainable Nuclear Industry Initiative (ESNII) aims at industrial application of fast reactor technology for a sustainable nuclear energy production. Currently four demonstration projects have a promising outlook in Europe, i.e. the ASTRID project in France, the MYRRHA project in Belgium, the ALFRED pan-European project to be realized in Romania, and SEALER in Sweden. Sodium and lead(-alloys) are envisaged as coolants for these reactors. Obviously, in the development of these reactors, thermal-hydraulics is recognized as a key challenge with emphasis on safety issues. This paper discusses the state-of-the-art knowledge with respect to experiments and simulation techniques as pursued in the Horizon 2020 SESAME (thermal hydraulics Simulations and Experiments for the Safety Assessment of MEtal cooled reactors) project.

Experience related to the safety of advanced LMFBR fuel elements

International Nuclear Information System (INIS)

Kerrisk, J.F.

1975-07-01

Experiments and experience relative to the safety of advanced fuel elements for the liquid metal fast breeder reactor are reviewed. The design and operating parameters and some of the unique features of advanced fuel elements are discussed breifly. Transient and steady state overpower operation and loss of sodium bond tests and experience are discussed in detail. Areas where information is lacking are also mentioned

Is the Experience of Thermal Pain Genetics Dependent?

Directory of Open Access Journals (Sweden)

Emilia Horjales-Araujo

2015-01-01

Full Text Available It is suggested that genetic variations explain a significant portion of the variability in pain perception; therefore, increased understanding of pain-related genetic influences may identify new targets for therapies and treatments. The relative contribution of the different genes to the variance in clinical and experimental pain responses remains unknown. It is suggested that the genetic contributions to pain perception vary across pain modalities. For example, it has been suggested that more than 60% of the variance in cold pressor responses can be explained by genetic factors; in comparison, only 26% of the variance in heat pain responses is explained by these variations. Thus, the selection of pain model might markedly influence the magnitude of the association between the pain phenotype and genetic variability. Thermal pain sensation is complex with multiple molecular and cellular mechanisms operating alone and in combination within the peripheral and central nervous system. It is thus highly probable that the thermal pain experience is affected by genetic variants in one or more of the pathways involved in the thermal pain signaling. This review aims to present and discuss some of the genetic variations that have previously been associated with different experimental thermal pain models.

Operational Experience from Solar Thermal Energy Projects

Science.gov (United States)

Cameron, C. P.

1984-01-01

Over the past few years, Sandia National Laboratories were involved in the design, construction, and operation of a number of DOE-sponsored solar thermal energy systems. Among the systems currently in operation are several industrial process heat projects and the Modular Industrial Solar Retrofit qualification test systems, all of which use parabolic troughs, and the Shenandoah Total Energy Project, which uses parabolic dishes. Operational experience has provided insight to both desirable and undesirable features of the designs of these systems. Features of these systems which are also relevant to the design of parabolic concentrator thermal electric systems are discussed. Other design features discussed are system control functions which were found to be especially convenient or effective, such as local concentrator controls, rainwash controls, and system response to changing isolation. Drive systems are also discussed with particular emphasis of the need for reliability and the usefulness of a manual drive capability.

Establishment of Safety Analysis System and Technology for CANDU Reactors

International Nuclear Information System (INIS)

Min, Byung Joo; Kim, W. Y.; Kim, H. T.; Rhee, B. W.; Yoon, C.; Kang, H. S.; Yoo, K. J.

2005-03-01

To improve the CANDU design/operation safety analysis codes and the CANDU safety analysis methodology, the following works have been done. From the development of the lattice codes (WIMS/CANDU), the lattice model simulates the real core lattice geometry and the effect of the pressure tube creep to the core lattice parameter has been evaluated. From the development of the 3-dimensional thermal-hydraulic analysis model of the moderator behavior (CFX4-CAMO), validation of the model against STERN Lab experiment has been executed. The butterfly-shaped grid structure and the 3-dimensional flow resistance model for porous media were developed and applied to the moderator analysis for Wolsong units 2/3/4. The single fuel channel analysis codes for blowdown and post-blowdown were unified by CATHENA. The 3-dimensional fuel channel analysis model (CFX-CACH) has been developed for validation of CATHENA fuel channel analysis model. The interlinking analysis system (CANVAS) of the thermal-hydraulic safety analysis codes for the primary heat transport system and containment system has been executed. The database system of core physics and thermal-hydraulics experimental data for safety analysis has been established on the URL: http://CANTHIS.kaeri.re.kr. For documentation and Standardization of the general safety analysis procedure, the general safety analysis procedure is developed and applied to a large break LOCA. The present research results can be utilized for establishment of the independent safety analysis technology and acquisition of the optimal safety analysis technology

Activation experiment for concrete blocks using thermal neutrons

Science.gov (United States)

Okuno, Koichi; Tanaka, Seiichiro

2017-09-01

Activation experiments for ordinary concrete, colemanite-peridotite concrete, B4C-loaded concrete, and limestone concrete are carried out using thermal neutrons. The results reveal that the effective dose for gamma rays from activated nuclides of colemanite-peridotite concrete is lower than that for the other types of concrete. Therefore, colemanite-peridotite concrete is useful for reducing radiation exposure for workers.

The experience of WNTI with Safety and Security Worldwide

International Nuclear Information System (INIS)

Neau, Henry-Jacques

2016-01-01

This paper gives an industrial perspective on safety and security issues based on the experience of WNTI members. It describes how safety is invested primarily in the package; not how the package is transported. Transport safety is therefore an engineering challenge, and all necessary technical information is available to enable this to be met. Security in transport involves various measures to guard against intentional malicious acts. The paper describes the international instruments relevant to security in the transport of nuclear fuel cycle materials and how both safety and security regulations must be coordinated and simplified to avoid conflicts. It considers potential risks, which must not be underestimated. However, the assessment of risks must be realistic and quantified, and the requirements placed on the industry appropriate. It is important to dispel exaggerated perceptions of danger in the minds of the public, politicians and regulators. (author)

Safety-oriented LWR research. Annual report 1987

International Nuclear Information System (INIS)

1988-06-01

The annual report 1987 includes the results of out-of-pile bundle experiments for severe fuel damage investigations (CORA), aerosol behaviour under core meltdown accident condition, dynamic behaviour of PWR containments, theoretical and experimental investigations of crack growth under thermal and thermomechanical fatigue loading, the thermal hydraulic code COMMIX-1B, the interfacial exchange processes in two-phase flow (NOVA-Program), the retention of penetrating iodine species, the development of exhaust filters and HEPA filter systems, advanced PWR's (APWR's) and related safety considerations, the HERA test facility and the KRISTA-program, and RELAP5/MOD2 post-test analysis of a forced feed reflood experiment. 20 papers are separately indexed in the database. (DG)

Leadership for safety: industrial experience.

Science.gov (United States)

Flin, R; Yule, S

2004-12-01

The importance of leadership for effective safety management has been the focus of research attention in industry for a number of years, especially in energy and manufacturing sectors. In contrast, very little research into leadership and safety has been carried out in medical settings. A selective review of the industrial safety literature for leadership research with possible application in health care was undertaken. Emerging findings show the importance of participative, transformational styles for safety performance at all levels of management. Transactional styles with attention to monitoring and reinforcement of workers' safety behaviours have been shown to be effective at the supervisory level. Middle managers need to be involved in safety and foster open communication, while ensuring compliance with safety systems. They should allow supervisors a degree of autonomy for safety initiatives. Senior managers have a prime influence on the organisation's safety culture. They need to continuously demonstrate a visible commitment to safety, best indicated by the time they devote to safety matters.

Some Findings from Thermal-Hydraulic Validation Tests for SMART Passive Safety System

Energy Technology Data Exchange (ETDEWEB)

Park, Hyun Sik; Bae, Hwang; Ryu, Sung-Uk; Ryu, Hyobong; Shin, Yong-Cheol; Min, Kyoung-Ho; Yi, Sung-Jae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

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, some findings from the validation tests for the SMART PSS 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 SDA revision and construction licensing. A test program to validate the performance of SMARS PSS was launched with an additional scaleddown test facility of SMART PSS, which will be installed at the existing SMART-ITL facility. In this paper, some findings from the validation tests of the SMART passive safety system during 2013-2014 were summarized. They include a couple of SMART PSS tests using active pumps and several 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.

Thermal characteristics during hydrogen fueling process of type IV cylinder

Energy Technology Data Exchange (ETDEWEB)

Kim, Sung Chan [Department of Fire and Disaster Prevention, Kyungil University, 33, Buhori, Hayang, Kyungsan 712-701 (Korea); Lee, Seung Hoon; Yoon, Kee Bong [Department of Mechanical Engineering, Chung Ang University, 221, Huksuk, Dongjak, Seoul 156-756 (Korea)

2010-07-15

Temperature increase during hydrogen fueling process is a significant safety concern of a high pressure hydrogen vessel. Hence, thermal characteristics of a Type IV cylinder during hydrogen filling process need to be understood. In this study, a series of experiments were conducted to quantify the temperature change of the cylinder during hydrogen filling to 35 MPa. Computational fluid dynamics (CFD) analysis was also conducted to simulate the conditions of the experiments. The results predicted by the CFD analysis show reasonable agreement with the experiments and the discrepancy between the CFD results and experimental results decrease with higher initial gas pressures. The upper and the lower parts of the vessel showed a temperature difference in the vertical direction. The upper gas temperature was higher than that of the lower part due to the buoyancy effect in the vessel. The maximum gas temperature was higher than the maximum temperature allowed in the ISO safety code (85 C) for the case in which the vessel was pressurized from 0 MPa to 35 MPa. This work contributes to the understanding of the thermal flow characteristics of the hydrogen filling process and notes that additional efforts should be made to guarantee the safety of a type IV cylinder during the hydrogen fueling process. (author)

CSRL-V ENDF/B-V 227-group neutron cross-section library and its application to thermal-reactor and criticality safety benchmarks

International Nuclear Information System (INIS)

Ford, W.E. III; Diggs, B.R.; Knight, J.R.; Greene, N.M.; Petrie, L.M.; Webster, C.C.; Westfall, R.M.; Wright, R.Q.; Williams, M.L.

1982-01-01

Characteristics and contents of the CSRL-V (Criticality Safety Reference Library based on ENDF/B-V data) 227-neutron-group AMPX master and pointwise cross-section libraries are described. Results obtained in using CSRL-V to calculate performance parameters of selected thermal reactor and criticality safety benchmarks are discussed

Pretest thermal analysis of the Tuff Water Migration/In-Situ Heater Experiment

International Nuclear Information System (INIS)

Bulmer, B.M.

1980-02-01

This report describes the pretest thermal analysis for the Tuff Water Migration/In-Situ Heater Experiment to be conducted in welded tuff in G-tunnel, Nevada Test Site. The parametric thermal modeling considers variable boiling temperature, tuff thermal conductivity, tuff emissivity, and heater operating power. For nominal tuff properties, some near field boiling is predicted for realistic operating power. However, the extent of boiling will be strongly determined by the ambient (100% water saturated) rock thermal conductivity. In addition, the thermal response of the heater and of the tuff within the dry-out zone (i.e., bounded by boiling isotherm) is dependent on the temperature variation of rock conductivity as well as the extent of induced boiling

Safety aspects and operating experience of LWR plants in Japan

International Nuclear Information System (INIS)

Aoki, S.; Hinoki, M.

1977-01-01

From the outset of nuclear power development in Japan, major emphasis has been placed on the safety of the nuclear power plants. There are now twelve nuclear power plants in operation with a total output of 6600 MWe. Their operating records were generally satisfactory, but in the 1974 to 1975 period, they experienced somewhat declined availability due to the repair work under the specific circumstances. After investigation of causes of troubles and the countermeasures thereof were made to ensure safety, they are now keeping good performance. In Japan, nuclear power plants are strictly subject to sufficient and careful inspection in compliance with the safety regulation, and are placed under stringent radiation control of employees. Under the various circumstances, however, the period of annual inspection tends to be prolonged more than originally planned, and this consequently is considered to be one of the causes of reduced availability. In order to develop nuclear power generation for the future, it is necessary to put further emphasis on the assurance of safety and to endeavor to devise measures to improve availability of the plants, based on the careful analysis of causes which reduce plant availability. This paper discusses the results of studies made for the following items from such viewpoints: (1) Safety and Operating Experience of LWR Nuclear Power Plants in Japan; a) Operating experience with light water reactors b) Improvements in design of light water reactors during the past ten years c) Analysis of the factors which affect plant availability; 2) Assurance of Safety and Measures to Increase Availability a) Measures for safety and environmental protection b) Measures to reduce radiation exposure of employees c) Appropriateness of maintenance and inspection work d) Measures to increase plant availability e) Measures to improve reliability of equipments and components; and 3) Future Technical Problems

Over one thousand patients with early stage endometriosis treated with the Helica Thermal Coagulator (HELICA): safety aspects.

Science.gov (United States)

Hill, Nicholas; McQueen, John; Morey, Richard; Hanna, Leila; Chandakas, Stefanos; El-Toukhy, Tarek; Erian, John

2006-07-01

To assess the safety of the Helica Thermal Coagulator in the laparoscopic treatment of early stage endometriosis. Retrospective, observational. The Princess Royal University Hospital, The Sloane and Chelsfield Park Hospitals, Kent, UK. One thousand and sixty patients with early stage endometriosis. All patients were treated laparoscopically with the Helica Thermal Coagulator; a new laparoscopic device that combines electrical energy with helium for the treatment of endometriosis. No major bladder, ureteric or bowel injuries occurred. The only complication was a perforated vagina from the cutting probe during dissection of the cul-de-sac in a patient with a vaginal endometriotic nodule. The Helica Thermal Coagulator is a safe device for the laparoscopic treatment of endometriosis.

PMK-2, the First Integral Thermal-Hydraulics Tests for the Safety Evaluation of VVER-440/213 Nuclear Power Plants

Directory of Open Access Journals (Sweden)

Gy. Ézsöl

2012-01-01

Full Text Available The PMK-2 facility is a full-pressure thermal-hydraulic model of the primary and partly the secondary circuit of the VVER-type units of Paks NPP. The facility was the first integral-type facility for VVERs. The PMK-2 was followed later by the PACTEL (for VVER-440, the ISB, and PSB for VVER-1000. Since the startup of the facility in 1985, 55 experiments have been performed primarily in international frameworks with the participation of experts from 29 European and overseas countries forming a scientific school to better understand VVER system behaviour and reach a high level of modelling of accident sequences. The ATHLET, CATHARE, and RELAP5 codes have been validated including both qualitative and quantitative assessments. The former was almost exclusively applied to the early phase of validation by integral experiments, while the quantitative assessments have been performed by the Fast Fourier Transform Based Method. Paper gives comprehensive information on the design features of PMK-2 facility with a special respect to the representativeness of phenomena, the experiments performed, and the results of the validation of ATHLET, CATHARE, and RELAP5 codes. Safety significance of the PMK-2 projects is also discussed.

PMK-2, the First Integral Thermal-Hydraulics Tests for the Safety Evaluation of VVER-440/213 Nuclear Power Plants

International Nuclear Information System (INIS)

Ezsol, G.; Perneczky, L.; Szabados, L.; Toth, I.

2012-01-01

The PMK-2 facility is a full-pressure thermal-hydraulic model of the primary and partly the secondary circuit of the VVER-type units of Paks NPP. The facility was the first integral-type facility for VVERs. The PMK-2 was followed later by the PACTEL (for VVER-440), the ISB, and PSB for VVER-1000. Since the startup of the facility in 1985, 55 experiments have been performed primarily in international frameworks with the participation of experts from 29 European and overseas countries forming a scientific school to better understand VVER system behaviour and reach a high level of modelling of accident sequences. The ATHLET, CATHARE, and RELAP5 codes have been validated including both qualitative and quantitative assessments. The former was almost exclusively applied to the early phase of validation by integral experiments, while the quantitative assessments have been performed by the Fast Fourier Transform Based Method. Paper gives comprehensive information on the design features of PMK-2 facility with a special respect to the representativeness of phenomena, the experiments performed, and the results of the validation of ATHLET, CATHARE, and RELAP5 codes. Safety significance of the PMK-2 projects is also discussed

Quality and microbial safety evaluation of new isotonic beverages upon thermal treatments.

Science.gov (United States)

Gironés-Vilaplana, Amadeo; Huertas, Juan-Pablo; Moreno, Diego A; Periago, Paula M; García-Viguera, Cristina

2016-03-01

In the present study, it was evaluated how two different thermal treatments (Mild and Severe) may affect the anthocyanin content, antioxidant capacity (ABTS(+), DPPH, and FRAP), quality (CIELAB colour parameters), and microbiological safety of a new isotonic drink made of lemon and maqui berry over a commercial storage simulation using a shelf life of 56days at two preservation temperature (7°C and 37°C). Both heat treatments did not affect drastically the anthocyanins content and their percentage of retention. The antioxidant capacity, probably because of the short time, was also not affected. The CIELAB colour parameters were affected by the heat, although the isotonic drinks remained with attractive red colour during shelf life. From a microbiological point of view, the Mild heat treatment with storage at 7°C is the ideal for the preservation of microbial growth, being useful for keeping the quality and safety of beverages in commercial life. Copyright © 2015. Published by Elsevier Ltd.

Experience of Milled Peat Burning at Thermal Electric Power Plant

Directory of Open Access Journals (Sweden)

G. I. Zhikhar

2013-01-01

Full Text Available The paper presents extensive knowledge and practical experience on burning of milled peat in the boilers of thermal electric power plants in Belarus and Russia. The accumulated experience can be used for solution of problems pertaining to substitution of some types of fuel imported to Belarus by milled peat which is extracted at many fuel effective peat enterprises of the Republic.

Experiments and analysis of thermal stresses around the nozzle of the reactor vessel

International Nuclear Information System (INIS)

Song, D.H.; Oh, J.H.; Song, H.K.; Park, D.S.; Shon, K.H.

1981-01-01

This report describes the results of analysis and experiments on the thermal stress around the reactor vessel nozzle performed to establish a capability of thermal stress analysis of pressure vessel subjected to thermal loadings. Firstly, heat conduction analysis during reactor design transients and analysis on the experimental model were performed using computer code FETEM-1 for the purpose of verification of FETEM-1 which was developed in 1979 and will be used to obtain the temperature distribution in a solid body under the steady-state and the transient conditions. The results of the analysis was compared to the results in the Stress Report of Kori-1 reactor vessel and those from experiments on the model, respectively

MCNP benchmark analyses of critical experiments for the Space Nuclear Thermal Propulsion program

International Nuclear Information System (INIS)

Selcow, E.C.; Cerbone, R.J.; Ludewig, H.; Mughabghab, S.F.; Schmidt, E.; Todosow, M.; Parma, E.J.; Ball, R.M.; Hoovler, G.S.

1993-01-01

Benchmark analyses have been performed of Particle Bed Reactor (PBR) critical experiments (CX) using the MCNP radiation transport code. The experiments have been conducted at the Sandia National Laboratory reactor facility in support of the Space Nuclear Thermal Propulsion (SNTP) program. The test reactor is a nineteen element water moderated and reflected thermal system. A series of integral experiments have been carried out to test the capabilities of the radiation transport codes to predict the performance of PBR systems. MCNP was selected as the preferred radiation analysis tool for the benchmark experiments. Comparison between experimental and calculational results indicate close agreement. This paper describes the analyses of benchmark experiments designed to quantify the accuracy of the MCNP radiation transport code for predicting the performance characteristics of PBR reactors

MCNP benchmark analyses of critical experiments for the Space Nuclear Thermal Propulsion program

Science.gov (United States)

Selcow, Elizabeth C.; Cerbone, Ralph J.; Ludewig, Hans; Mughabghab, Said F.; Schmidt, Eldon; Todosow, Michael; Parma, Edward J.; Ball, Russell M.; Hoovler, Gary S.

1993-01-01

Benchmark analyses have been performed of Particle Bed Reactor (PBR) critical experiments (CX) using the MCNP radiation transport code. The experiments have been conducted at the Sandia National Laboratory reactor facility in support of the Space Nuclear Thermal Propulsion (SNTP) program. The test reactor is a nineteen element water moderated and reflected thermal system. A series of integral experiments have been carried out to test the capabilities of the radiation transport codes to predict the performance of PBR systems. MCNP was selected as the preferred radiation analysis tool for the benchmark experiments. Comparison between experimental and calculational results indicate close agreement. This paper describes the analyses of benchmark experiments designed to quantify the accuracy of the MCNP radiation transport code for predicting the performance characteristics of PBR reactors.

Heat experiment design to estimate temperature dependent thermal properties

International Nuclear Information System (INIS)

Romanovski, M

2008-01-01

Experimental conditions are studied to optimize transient experiments for estimating temperature dependent thermal conductivity and volumetric heat capacity. A mathematical model of a specimen is the one-dimensional heat equation with boundary conditions of the second kind. Thermal properties are assumed to vary nonlinearly with temperature. Experimental conditions refer to the thermal loading scheme, sampling times and sensor location. A numerical model of experimental configurations is studied to elicit the optimal conditions. The numerical solution of the design problem is formulated on a regularization scheme with a stabilizer minimization without a regularization parameter. An explicit design criterion is used to reveal the optimal sensor location, heating duration and flux magnitude. Results obtained indicate that even the strongly nonlinear experimental design problem admits the aggregation of its solution and has a strictly defined optimal measurement scheme. Additional region of temperature measurements with allowable identification error is revealed.

The CERN Detector Safety System for LHC Experiments

CERN Document Server

Lüders, S; Morpurgo, G; Schmeling, S M

2003-01-01

The Detector Safety System (DSS), developed at CERN in common for the four LHC experiments under the auspices of the Joint Controls Project (JCOP), will be responsible for assuring the equipment protection for these experiments. Therefore, the DSS requires a high degree of both availability and reliability. It is composed of a Front-end and a Back-end part. The Front-end is based on a redundant Siemens PLC, to which the safety-critical part of the DSS task is delegated. The PLC Front-end is capable of running autonomously and of automati-cally taking predefined protective actions whenever re-quired. It is supervised and configured by the CERN-cho-sen PVSS SCADA system via a Siemens OPC server. The supervisory layer provides the operator with a status display and with limited online reconfiguration capabili-ties. Configuration of the code running in the PLCs is completely data driven via the contents of a ?Configura-tion Database?. Thus, the DSS can easily adapt to the different and constantly evolving require...

International Handbook of Evaluated Criticality Safety Benchmark Experiments - ICSBEP (DVD), Version 2013

International Nuclear Information System (INIS)

2013-01-01

The Criticality Safety Benchmark Evaluation Project (CSBEP) was initiated in October of 1992 by the United States Department of Energy. The project quickly became an international effort as scientists from other interested countries became involved. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) became an official activity of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) in 1995. This handbook contains criticality safety benchmark specifications that have been derived from experiments performed at various nuclear critical experiment facilities around the world. The benchmark specifications are intended for use by criticality safety engineers to validate calculational techniques used to establish minimum subcritical margins for operations with fissile material and to determine criticality alarm requirement and placement. Many of the specifications are also useful for nuclear data testing. Example calculations are presented; however, these calculations do not constitute a validation of the codes or cross section data. The evaluated criticality safety benchmark data are given in nine volumes. These volumes span nearly 66,000 pages and contain 558 evaluations with benchmark specifications for 4,798 critical, near critical or subcritical configurations, 24 criticality alarm placement/shielding configurations with multiple dose points for each and 200 configurations that have been categorised as fundamental physics measurements that are relevant to criticality safety applications. New to the Handbook are benchmark specifications for Critical, Bare, HEU(93.2)- Metal Sphere experiments referred to as ORSphere that were performed by a team of experimenters at Oak Ridge National Laboratory in the early 1970's. A photograph of this assembly is shown on the front cover

Operating experiences since rise-to-power test in high temperature engineering test reactor (HTTR)

International Nuclear Information System (INIS)

Tochio, Daisuke; Watanabe, Shuji; Motegi, Toshihiro; Kawano, Shuichi; Kameyama, Yasuhiko; Sekita, Kenji; Kawasaki, Kozo

2007-03-01

The rise-to-power test of the High Temperature Engineering Test Reactor (HTTR) was actually started in April 2000. The rated thermal power of 30MW and the rated reactor outlet coolant temperature of 850degC were achieved in the middle of Dec. 2001. After that, the reactor thermal power of 30MW and the reactor outlet coolant temperature of 950degC were achieved in the final rise-to-power test in April 2004. After receiving the operation licensing at 850degC, the safety demonstration tests have conducted to demonstrate inherent safety features of the HTGRs as well as to obtain the core and plant transient data for validation of safety analysis codes and for establishment of safety design and evaluation technologies. This paper summarizes the HTTR operating experiences for six years from start of the rise-to-power test that are categorized into (1) Operating experiences related to advanced gas-cooled reactor design, (2) Operating experiences for improvement of the performance, (3) Operating experiences due to fail of system and components. (author)

PPOOLEX experiments on thermal stratification and mixing

Energy Technology Data Exchange (ETDEWEB)

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

2009-08-15

The results of the thermal stratification experiments in 2008 with the PPOOLEX test facility are presented. PPOOLEX is a closed vessel divided into two compartments, dry well and wet well. Extra temperature measurements for capturing different aspects of the investigated phenomena were added before the experiments. The main purpose of the experiment series was to generate verification data for evaluating the capability of GOTHIC code to predict stratification and mixing phenomena. Altogether six experiments were carried out. Heat-up periods of several thousand seconds by steam injection into the dry well compartment and from there into the wet well water pool were recorded. The initial water bulk temperature was 20 deg. C. Cooling periods of several days were included in three experiments. A large difference between the pool bottom and top layer temperature was measured when small steam flow rates were used. With higher flow rates the mixing effect of steam discharge delayed the start of stratification until the pool bulk temperature exceeded 50 deg. C. The stratification process was also different in these two cases. With a small flow rate stratification was observed only above and just below the blowdown pipe outlet elevation. With a higher flow rate over a 30 deg. C temperature difference between the pool bottom and pipe outlet elevation was measured. Elevations above the pipe outlet indicated almost linear rise until the end of steam discharge. During the cooling periods the measurements of the bottom third of the pool first had an increasing trend although there was no heat input from outside. This was due to thermal diffusion downwards from the higher elevations. Heat-up in the gas space of the wet well was quite strong, first due to compression by pressure build-up and then by heat conduction from the hot dry well compartment via the intermediate floor and test vessel walls and by convection from the upper layers of the hot pool water. The gas space

PPOOLEX experiments on thermal stratification and mixing

International Nuclear Information System (INIS)

Puustinen, M.; Laine, J.; Raesaenen, A.

2009-08-01

The results of the thermal stratification experiments in 2008 with the PPOOLEX test facility are presented. PPOOLEX is a closed vessel divided into two compartments, dry well and wet well. Extra temperature measurements for capturing different aspects of the investigated phenomena were added before the experiments. The main purpose of the experiment series was to generate verification data for evaluating the capability of GOTHIC code to predict stratification and mixing phenomena. Altogether six experiments were carried out. Heat-up periods of several thousand seconds by steam injection into the dry well compartment and from there into the wet well water pool were recorded. The initial water bulk temperature was 20 deg. C. Cooling periods of several days were included in three experiments. A large difference between the pool bottom and top layer temperature was measured when small steam flow rates were used. With higher flow rates the mixing effect of steam discharge delayed the start of stratification until the pool bulk temperature exceeded 50 deg. C. The stratification process was also different in these two cases. With a small flow rate stratification was observed only above and just below the blowdown pipe outlet elevation. With a higher flow rate over a 30 deg. C temperature difference between the pool bottom and pipe outlet elevation was measured. Elevations above the pipe outlet indicated almost linear rise until the end of steam discharge. During the cooling periods the measurements of the bottom third of the pool first had an increasing trend although there was no heat input from outside. This was due to thermal diffusion downwards from the higher elevations. Heat-up in the gas space of the wet well was quite strong, first due to compression by pressure build-up and then by heat conduction from the hot dry well compartment via the intermediate floor and test vessel walls and by convection from the upper layers of the hot pool water. The gas space

Experience With Laser Safety In The USA--A Review

Science.gov (United States)

Sliney, David H.

1986-10-01

Following several research programs in the 1960's aimed at studying the adverse biological effects of lasers and other optical radiation sources, laser occupational exposure limits were set and general safety standards were developed. Today, the experience from laser accidents and the development of new lasers and new applications have altered the format of the exposure limits and the safety procedures. It is critically important to distinguish between different biological injury mechanisms. The biological effects of ultraviolet radiation upon the skin and eye are additive over a period of at least one workday, and require different safety procedures. The scattered UV irradiance from excimer lasers may be quite hazardous, depending upon wavelength and action spectra. Since laser technology is young, the exposure of an individual in natural sunlight must be studied to evaluate the potential for chronic effects. The safety measures necessary in the use of lasers depend upon a hazard evaluation. The appropriate control measures and alternate means of enclosure, baffling, and operational control measures are presented. Present laser safety standards are explained briefly. Eye protective techniques and eyewear are considered for a variety of sources. The optical properties of enclosure materials are also discussed.

Large-Scale Spacecraft Fire Safety Experiments in ISS Resupply Vehicles

Science.gov (United States)

Ruff, Gary A.; Urban, David

2013-01-01

Our understanding of the fire safety risk in manned spacecraft has been limited by the small scale of the testing we have been able to conduct in low-gravity. Fire growth and spread cannot be expected to scale linearly with sample size so we cannot make accurate predictions of the behavior of realistic scale fires in spacecraft based on the limited low-g testing to date. As a result, spacecraft fire safety protocols are necessarily very conservative and costly. Future crewed missions are expected to be longer in duration than previous exploration missions outside of low-earth orbit and accordingly, more complex in terms of operations, logistics, and safety. This will increase the challenge of ensuring a fire-safe environment for the crew throughout the mission. Based on our fundamental uncertainty of the behavior of fires in low-gravity, the need for realistic scale testing at reduced gravity has been demonstrated. To address this concern, a spacecraft fire safety research project is underway to reduce the uncertainty and risk in the design of spacecraft fire safety systems by testing at nearly full scale in low-gravity. This project is supported by the NASA Advanced Exploration Systems Program Office in the Human Exploration and Operations Mission Directorate. The activity of this project is supported by an international topical team of fire experts from other space agencies to maximize the utility of the data and to ensure the widest possible scrutiny of the concept. The large-scale space flight experiment will be conducted on three missions; each in an Orbital Sciences Corporation Cygnus vehicle after it has deberthed from the ISS. Although the experiment will need to meet rigorous safety requirements to ensure the carrier vehicle does not sustain damage, the absence of a crew allows the fire products to be released into the cabin. The tests will be fully automated with the data downlinked at the conclusion of the test before the Cygnus vehicle reenters the

Modernization of Unit 2 at Oskarshamn NPP- Main Objectives, Experience from Design, Separation of Operational and Nuclear Safety Equipment - Lessons Learned

International Nuclear Information System (INIS)

Kanaan, Salah K.

2015-01-01

This paper aims to give a picture of Oskarshamn Nuclear Power Plant (OKG) experience from design for one of the biggest modernization project in the world and focuses on what was learned that is specific to robustness of electrical power systems, especially through Fukushima Station Blackout (SBO). The planning for unit 2 at OKG was initiated in 1967 and the plant was completed on time and was synchronized to the grid October 2, 1974 and is of type BWR. Unit 2 was originally on 580 MW. In 1982 a thermal power up-rate was performed, from 1700 MWh to 1800 MWh (106% reactor output). A decision was made to perform a modernization and a new power up-rate to 850 MW and there were several reasons for this decision; New safety regulations from Swedish Radiation Safety Authority (SSM), Ageing of important components and the initial focus was on safety and availability - Project Plant Life Extension (Plex) was established and became the largest nuclear power modernization in the world. The modernization will lead to: - New safety concept with 4 divisions instead for existing 2 with 2 new buildings South Electrical Building (SEB) and North Electrical Building (NEB); - Completely new software - based equipments for monitoring, control and I and C; - New Low Pressure Turbine, new generator and main transformer; - New MCR and simulator; - Compliance with modern reactor safety requirements; - Redundancy, Separation, Diversification, Earthquake; - Reinforcement of existing safety functions; - New Electricity - I and C (electric power incl. reinforced emergency power and control systems); - New buildings for Electricity - I and C; - Reinforcement of existing process systems as well as installation of new ones. Based on studies and good experiences on how to separate the operational and the safety equipment, the project led to a completely new safety concept. The safety concept is based on fully understanding the safety system that shall encompass all of the elements required to

Energy Storage Thermal Safety | Transportation Research | NREL

Science.gov (United States)

reaction/thermal runaway, internal short circuit, and electrical/chemical/thermal network models are used contributions to the U.S. Department of Energy's Computer-Aided Engineering of Batteries (CAEBAT) project Li-ion battery geometries. Chemical components in Li-ion batteries become thermally unstable when

Review of studies on criticality safety evaluation and criticality experiment methods

International Nuclear Information System (INIS)

Naito, Yoshitaka; Yamamoto, Toshihiro; Misawa, Tsuyoshi; Yamane, Yuichi

2013-01-01

Since the early 1960s, many studies on criticality safety evaluation have been conducted in Japan. Computer code systems were developed initially by employing finite difference methods, and more recently by using Monte Carlo methods. Criticality experiments have also been carried out in many laboratories in Japan as well as overseas. By effectively using these study results, the Japanese Criticality Safety Handbook was published in 1988, almost the intermediate point of the last 50 years. An increased interest has been shown in criticality safety studies, and a Working Party on Nuclear Criticality Safety (WPNCS) was set up by the Nuclear Science Committee of Organisation Economic Co-operation and Development in 1997. WPNCS has several task forces in charge of each of the International Criticality Safety Benchmark Evaluation Program (ICSBEP), Subcritical Measurement, Experimental Needs, Burn-up Credit Studies and Minimum Critical Values. Criticality safety studies in Japan have been carried out in cooperation with WPNCS. This paper describes criticality safety study activities in Japan along with the contents of the Japanese Criticality Safety Handbook and the tasks of WPNCS. (author)

New safety experiments in decommissioned superheated steam reactor at Karlstein

International Nuclear Information System (INIS)

Koerting, K.

1986-01-01

This article gives a concise summary of the Status Report of the Superheated Steam Reactor Safety Program (PHDR) Project, held at KfK on Dec. 5, 1985. The results discussed dealt with fire experiments, shock tests simulating airplane crashes, temperature shocks in the reactor pressure vessel, studies of crack detection in pressure vessels and blasting experiments associated with nuclear plant decommissioning

Numerical experiments on thermal convection of highly compressible fluids with variable viscosity and thermal conductivity: Implications for mantle convection of super-Earths

Science.gov (United States)

Kameyama, Masanori; Yamamoto, Mayumi

2018-01-01

We conduct a series of numerical experiments of thermal convection of highly compressible fluids in a two-dimensional rectangular box, in order to study the mantle convection on super-Earths. The thermal conductivity and viscosity are assumed to exponentially depend on depth and temperature, respectively, while the variations in thermodynamic properties (thermal expansivity and reference density) with depth are taken to be relevant for the super-Earths with 10 times the Earth's. From our experiments we identified a distinct regime of convecting flow patterns induced by the interplay between the adiabatic temperature change and the spatial variations in viscosity and thermal conductivity. That is, for the cases with strong temperature-dependent viscosity and depth-dependent thermal conductivity, a "deep stratosphere" of stable thermal stratification is formed at the base of the mantle, in addition to thick stagnant lids at their top surfaces. In the "deep stratosphere", the fluid motion is insignificant particularly in the vertical direction in spite of smallest viscosity owing to its strong dependence on temperature. Our finding may further imply that some of super-Earths which are lacking in mobile tectonic plates on their top surfaces may have "deep stratospheres" at the base of their mantles.

Review of irradiation experiments for water reactor safety research

International Nuclear Information System (INIS)

Tobioka, Toshiaki

1977-02-01

A review is made of irradiation experiments for water reactor safety research under way in both commercial power plants and test reactors. Such experiments are grouped in two; first, LWR fuel performance under normal and abnormal operating conditions, and second, irradiation effects on fracture toughness in LWR vessels. In the former are fuel densification, swelling, and the influence of power ramp and cycling on fuel rod, and also fuel rod behavior under accident conditions in in-reactor experiment. In the latter are the effects of neutron exposure level on the ferritic steel of pressure vessels, etc.. (auth.)

Data report of BWR post-CHF tests. Transient core thermal-hydraulic test program. Contract research

International Nuclear Information System (INIS)

Iguchi, Tadashi; Itoh, Hideo; Kiuchi, Toshio; Watanabe, Hironori; Kimura, Mamoru; Anoda, Yoshinari

2001-03-01

JAERI has been performing transient core thermal-hydraulic test program. In the program, authors performed BWR/ABWR DBE simulation tests with a test facility, which can simulate BWR/ABWR transients. The test facility has a 4 x 4 bundle core simulator with 15-rod heaters and one non-heated rod. Through the tests, authors quantified the thermal safety margin for core cooling. In order to quantify the thermal safety margin, authors collected experimental data on post-CHF. The data are essential for the evaluation of clad temperature transient when core heat-up occurs during DBEs. In comparison with previous post-CHF tests, present experiments were performed in much wider experimental condition, covering high clad temperature, low to high pressure and low to high mass flux. Further, data at wider elevation (lower to higher elevation of core) were obtained in the present experiments, which make possible to discuss the effect of axial position on thermal-hydraulics, while previous works usually discuss the thermal-hydraulics at the position where the first heat-up occurs. This data report describes test procedure, test condition and major experimental data of post-CHF tests. (author)

The Detector Safety System of the ATLAS experiment

International Nuclear Information System (INIS)

Beltramello, O; Burckhart, H J; Franz, S; Jaekel, M; Jeckel, M; Lueders, S; Morpurgo, G; Santos Pedrosa, F dos; Pommes, K; Sandaker, H

2009-01-01

The ATLAS detector at the Large Hadron Collider at CERN is one of the most advanced detectors for High Energy Physics experiments ever built. It consists of the order of ten functionally independent sub-detectors, which all have dedicated services like power, cooling, gas supply. A Detector Safety System has been built to detect possible operational problems and abnormal and potentially dangerous situations at an early stage and, if needed, to bring the relevant part of ATLAS automatically into a safe state. The procedures and the configuration specific to ATLAS are described in detail and first operational experience is given.

Hydrogen safety

International Nuclear Information System (INIS)

Frazier, W.R.

1991-01-01

The NASA experience with hydrogen began in the 1950s when the National Advisory Committee on Aeronautics (NACA) research on rocket fuels was inherited by the newly formed National Aeronautics and Space Administration (NASA). Initial emphasis on the use of hydrogen as a fuel for high-altitude probes, satellites, and aircraft limited the available data on hydrogen hazards to small quantities of hydrogen. NASA began to use hydrogen as the principal liquid propellant for launch vehicles and quickly determined the need for hydrogen safety documentation to support design and operational requirements. The resulting NASA approach to hydrogen safety requires a joint effort by design and safety engineering to address hydrogen hazards and develop procedures for safe operation of equipment and facilities. NASA also determined the need for rigorous training and certification programs for personnel involved with hydrogen use. NASA's current use of hydrogen is mainly for large heavy-lift vehicle propulsion, which necessitates storage of large quantities for fueling space shots and for testing. Future use will involve new applications such as thermal imaging

The safety experience of New Zealand adventure tourism operators.

Science.gov (United States)

Bentley, Tim A; Page, Stephen; Walker, Linda

2004-01-01

This survey examined parameters of the New Zealand adventure tourism industry client injury risk. The research also sought to establish priorities for intervention to reduce adventure tourism risk, and identify client injury control measures currently in place (or absent) in the New Zealand adventure tourism industry, with a view to establishing guidelines for the development of effective adventure tourism safety management systems. This 2003 survey builds upon an exploratory study of New Zealand adventure tourism safety conducted by us during 1999. A postal questionnaire was used to survey all identifiable New Zealand adventure tourism operators. The questionnaire asked respondents about their recorded client injury experience, perceptions of client injury risk factors, safety management practices, and barriers to safety. Some 27 adventure tourism activities were represented among the responding sample (n=96). The highest client injury risk was reported in the snow sports, bungee jumping and horse riding sectors, although serious underreporting of minor injuries was evident across the industry. Slips, trips and falls (STF) were the major client injury mechanisms, and a range of risk factors for client injuries were identified. Safety management measures were inconsistently applied across the industry. The industry should consider the implications of poor injury reporting standards and safety management practices generally. Specifically, the industry should consider risk management that focuses on minor (e.g., STF) as well as catastrophic events.

Development of realistic thermal hydraulic system analysis code

Energy Technology Data Exchange (ETDEWEB)

Lee, Won Jae; Chung, B. D; Kim, K. D. [and others

2002-05-01

The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others.

Development of realistic thermal hydraulic system analysis code

International Nuclear Information System (INIS)

Lee, Won Jae; Chung, B. D; Kim, K. D.

2002-05-01

The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others

International benchmark study of advanced thermal hydraulic safety analysis codes against measurements on IEA-R1 research reactor

Energy Technology Data Exchange (ETDEWEB)

Hainoun, A., E-mail: pscientific2@aec.org.sy [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Doval, A. [Nuclear Engineering Department, Av. Cmdt. Luis Piedrabuena 4950, C.P. 8400 S.C de Bariloche, Rio Negro (Argentina); Umbehaun, P. [Centro de Engenharia Nuclear – CEN, IPEN-CNEN/SP, Av. Lineu Prestes 2242-Cidade Universitaria, CEP-05508-000 São Paulo, SP (Brazil); Chatzidakis, S. [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States); Ghazi, N. [Atomic Energy Commission of Syria (AECS), Nuclear Engineering Department, P.O. Box 6091, Damascus (Syrian Arab Republic); Park, S. [Research Reactor Design and Engineering Division, Basic Science Project Operation Dept., Korea Atomic Energy Research Institute (Korea, Republic of); Mladin, M. [Institute for Nuclear Research, Campului Street No. 1, P.O. Box 78, 115400 Mioveni, Arges (Romania); Shokr, A. [Division of Nuclear Installation Safety, Research Reactor Safety Section, International Atomic Energy Agency, A-1400 Vienna (Austria)

2014-12-15

Highlights: • A set of advanced system thermal hydraulic codes are benchmarked against IFA of IEA-R1. • Comparative safety analysis of IEA-R1 reactor during LOFA by 7 working teams. • This work covers both experimental and calculation effort and presents new out findings on TH of RR that have not been reported before. • LOFA results discrepancies from 7% to 20% for coolant and peak clad temperatures are predicted conservatively. - Abstract: In the framework of the IAEA Coordination Research Project on “Innovative methods in research reactor analysis: Benchmark against experimental data on neutronics and thermal hydraulic computational methods and tools for operation and safety analysis of research reactors” the Brazilian research reactor IEA-R1 has been selected as reference facility to perform benchmark calculations for a set of thermal hydraulic codes being widely used by international teams in the field of research reactor (RR) deterministic safety analysis. The goal of the conducted benchmark is to demonstrate the application of innovative reactor analysis tools in the research reactor community, validation of the applied codes and application of the validated codes to perform comprehensive safety analysis of RR. The IEA-R1 is equipped with an Instrumented Fuel Assembly (IFA) which provided measurements for normal operation and loss of flow transient. The measurements comprised coolant and cladding temperatures, reactor power and flow rate. Temperatures are measured at three different radial and axial positions of IFA summing up to 12 measuring points in addition to the coolant inlet and outlet temperatures. The considered benchmark deals with the loss of reactor flow and the subsequent flow reversal from downward forced to upward natural circulation and presents therefore relevant phenomena for the RR safety analysis. The benchmark calculations were performed independently by the participating teams using different thermal hydraulic and safety

Thermal hydraulic issues and challenges for current and new generation FBRs

Energy Technology Data Exchange (ETDEWEB)

Chellapandi, P.; Velusamy, K., E-mail: kvelu@igcar.gov.in

2015-12-01

Highlights: • We present challenges in thermal hydraulic design of sodium cooled fast reactors. • We present roadmap of Indian fast reactor program and innovative design concepts. • Analysis methodology for thermal striping and thermal stratification are highlighted. • Design solutions for gas entrainment are presented. • Experimental approaches for normal and post accident decay heat removal are highlighted. - Abstract: Pool type sodium cooled fast reactors pose several design challenges and among them, certain thermal hydraulics and structural mechanics issues are special. High frequency temperature fluctuations due to thermal striping, thermal stratifications and sodium free level fluctuations at the liquid–cover gas interfaces are to be investigated carefully to eliminate high cycle thermal fatigue of structures. Solutions to address the core thermal hydraulics call for high power computing. Innovative concepts and methods are developed to carry out plant dynamics and safety studies. Particularly, extensive numerical and experimental simulation techniques are needed for understanding and solving the gas entrainment mechanisms and its effects on core safety. Though decay heat removal through natural convection is achievable in a pool type SFR, demonstration of design solutions conceived in the reactor and performance of diverse systems under all operating conditions, especially over prolonged station blackout situations needs advanced CFD computations and should be validated by relatively large scale simulated experiments. These issues are addressed in this paper under five broad topics: special thermal hydraulic issues to be addressed in SFR, thermal hydraulic design and analysis, plant dynamics studies, safety studies and evolving thermal hydraulic studies for the future FBRs. The 500 MWe Prototype Fast Breeder Reactor (PFBR) is taken as the reference design for addressing the issues. Indian fast reactor programme is highlighted in the introduction

Investigating the experiences of New Zealand MRI technologists: Exploring intra-orbital metallic foreign body safety practices

Energy Technology Data Exchange (ETDEWEB)

Jacobs, Philippa K [River Radiology, Victoria Clinic, 750 Victoria Street, Hamilton, Waikato (New Zealand); Henwood, Suzanne [Unitec - Medical Imaging, Unitec Ratanui Street Henderson, Auckland (New Zealand); River Radiology, Victoria Clinic, 750 Victoria Street, Hamilton, Waikato (New Zealand)

2013-12-15

Qualitative research is lacking regarding the experiences of magnetic resonance imaging (MRI) technologists and their involvement in workplace safety practices. This article provides a gateway to explore, describe and document experiences of MRI technologists in New Zealand (NZ) pertaining to intra-orbital metallic foreign body (IMFB) safety practices. This phenomenological study describes the experiences of seven MRI technologists all with a minimum of 5 years' NZ work experience in MRI. The MRI technologists were interviewed face-to-face regarding their professional IMFB workplace experiences in order to explore historical, current and potential issues. Findings demonstrated that aspects of organization and administration are fundamentally important to MRI technologists. Varying levels of education and knowledge, as well as experience and skills gained, have significantly impacted on MRI technologists’ level of confidence and control in IMFB practices. Participants’ descriptions of their experiences in practice regarding decision-making capabilities further highlight the complexity of these themes. A model was developed to demonstrate the interrelated nature of the themes and the complexity of the situation in totality. Findings of this study have provided insight into the experiences of MRI technologists pertaining to IMFB safety practices and highlighted inconsistencies. It is hoped that these findings will contribute to and improve the level of understanding of MRI technologists and the practices and protocols involved in IMFB safety screening. The scarcity of available literature regarding IMFB safety practices highlights that more research is required to investigate additional aspects that could improve MRI technologists’ experiences.

Investigating the experiences of New Zealand MRI technologists: Exploring intra-orbital metallic foreign body safety practices

International Nuclear Information System (INIS)

Jacobs, Philippa K; Henwood, Suzanne

2013-01-01

Qualitative research is lacking regarding the experiences of magnetic resonance imaging (MRI) technologists and their involvement in workplace safety practices. This article provides a gateway to explore, describe and document experiences of MRI technologists in New Zealand (NZ) pertaining to intra-orbital metallic foreign body (IMFB) safety practices. This phenomenological study describes the experiences of seven MRI technologists all with a minimum of 5 years' NZ work experience in MRI. The MRI technologists were interviewed face-to-face regarding their professional IMFB workplace experiences in order to explore historical, current and potential issues. Findings demonstrated that aspects of organization and administration are fundamentally important to MRI technologists. Varying levels of education and knowledge, as well as experience and skills gained, have significantly impacted on MRI technologists’ level of confidence and control in IMFB practices. Participants’ descriptions of their experiences in practice regarding decision-making capabilities further highlight the complexity of these themes. A model was developed to demonstrate the interrelated nature of the themes and the complexity of the situation in totality. Findings of this study have provided insight into the experiences of MRI technologists pertaining to IMFB safety practices and highlighted inconsistencies. It is hoped that these findings will contribute to and improve the level of understanding of MRI technologists and the practices and protocols involved in IMFB safety screening. The scarcity of available literature regarding IMFB safety practices highlights that more research is required to investigate additional aspects that could improve MRI technologists’ experiences

TITAN: an advanced three-dimensional neutronics/thermal-hydraulics code for light water reactor safety analysis

International Nuclear Information System (INIS)

Griggs, D.P.; Kazimi, M.S.; Henry, A.F.

1982-01-01

The initial development of TITAN, a three-dimensional coupled neutronics/thermal-hydraulics code for LWR safety analysis, has been completed. The transient neutronics code QUANDRY has been joined to the two-fluid thermal-hydraulics code THERMIT with the appropriate feedback mechanisms modeled. A detailed steady-state and transient coupling scheme based on the tandem technique was implemented in accordance with the important structural and operational characteristics of QUANDRY and THERMIT. A two channel sample problem formed the basis for steady-state and transient analyses performed with TITAN. TITAN steady-state results were compared with those obtained with MEKIN and showed good agreement. Null transients, simulated turbine trip transients, and a rod withdrawal transient were analyzed with TITAN and reasonable results were obtained

Experience with performance based training of nuclear criticality safety engineers

International Nuclear Information System (INIS)

Taylor, R.G.

1993-01-01

Historically, new entrants to the practice of nuclear criticality safety have learned their job primarily by on-the-job training (OJT) often by association with an experienced nuclear criticality safety engineer who probably also learned their job by OJT. Typically, the new entrant learned what he/she needed to know to solve a particular problem and accumulated experience as more problems were solved. It is likely that more formalism will be required in the future. Current US Department of Energy requirements for those positions which have to demonstrate qualification indicate that it should be achieved by using a systematic approach such as performance based training (PBT). Assuming that PBT would be an acceptable mechanism for nuclear criticality safety engineer training in a more formal environment, a site-specific analysis of the nuclear criticality safety engineer job was performed. Based on this analysis, classes are being developed and delivered to a target audience of newer nuclear criticality safety engineers. Because current interest is in developing training for selected aspects of the nuclear criticality safety engineer job, the analysis i's incompletely developed in some areas. Details of this analysis are provided in this report

Nurses' experiences and perspectives on medication safety practices: an explorative qualitative study

NARCIS (Netherlands)

Smeulers, Marian; Onderwater, Astrid T.; van Zwieten, Myra C. B.; Vermeulen, Hester

2014-01-01

To explore nurses' experiences with and perspectives on preventing medication administration errors. Insight into nurses' experiences with and perspectives on preventing medication administration errors is important and can be utilised to tailor and implement safety practices. A qualitative

Safety research programs sponsored by Office of Nuclear Regulatory Research. Quarterly progress report, July 1-September 30, 1983. Volume 3, No. 3

Energy Technology Data Exchange (ETDEWEB)

Weiss, A J [comp.

1984-01-01

The projects reported are the following: HTGR Safety Evaluation, SSC Development, Validation and Application, CRBR Balance of Plant Modeling, Thermal-Hydraulic Reactor Safety Experiments, LWR Plant Analyzer Development, LWR Code Assessment and Application, Thermal Reactor Code Development (RAMONA-3B); Stress Corrosion Cracking of PWR Steam Generator Tubing, Bolting Failure Analysis, Probability Based Load Combinations for Design of Category I Structures, Mechanical Piping Benchmark Problems; Human Error Data for Nuclear Power Plant Safety-Related Events, and Human Factors in Nuclear Power Plant Safeguards.

Enhanced thermal safety and high power performance of carbon-coated LiFePO4 olivine cathode for Li-ion batteries

Science.gov (United States)

Zaghib, K.; Dubé, J.; Dallaire, A.; Galoustov, K.; Guerfi, A.; Ramanathan, M.; Benmayza, A.; Prakash, J.; Mauger, A.; Julien, C. M.

2012-12-01

The carbon-coated LiFePO4 Li-ion oxide cathode was studied for its electrochemical, thermal, and safety performance. This electrode exhibited a reversible capacity corresponding to more than 89% of the theoretical capacity when cycled between 2.5 and 4.0 V. Cylindrical 18,650 cells with carbon-coated LiFePO4 also showed good capacity retention at higher discharge rates up to 5C rate with 99.3% coulombic efficiency, implying that the carbon coating improves the electronic conductivity. Hybrid Pulse Power Characterization (HPPC) test performed on LiFePO4 18,650 cell indicated the suitability of this carbon-coated LiFePO4 for high power HEV applications. The heat generation during charge and discharge at 0.5C rate, studied using an Isothermal Microcalorimeter (IMC), indicated cell temperature is maintained in near ambient conditions in the absence of external cooling. Thermal studies were also investigated by Differential Scanning Calorimeter (DSC) and Accelerating Rate Calorimeter (ARC), which showed that LiFePO4 is safer, upon thermal and electrochemical abuse, than the commonly used lithium metal oxide cathodes with layered and spinel structures. Safety tests, such as nail penetration and crush test, were performed on LiFePO4 and LiCoO2 cathode based cells, to investigate on the safety hazards of the cells upon severe physical abuse and damage.

Views on the future of thermal hydraulic modeling

Energy Technology Data Exchange (ETDEWEB)

Ishii, M. [Purdue Univ., West Lafayette, IN (United States)

1997-07-01

It is essential for the U.S. NRC to sustain the highest level of the thermal-hydraulics and reactor safety research expertise and continuously improve their accident analysis capability. Such expertise should span over four different areas which are strongly related to each other. These are: (1) Reactor Safety Code Development, (2) Two-phase Flow Modeling, (3) Instrumentation and Fundamental Experimental Research, and (4) Separate Effect and Integral Test. The NRC is already considering a new effort in the area of advanced thermal-hydraulics effort. Its success largely depends on the availability of a significantly improved two-phase flow formulation and constitutive relations supported by detailed experimental data. Therefore, it is recommended that the NRC start significant research efforts in the areas of two-phase flow modeling, instrumentation, basic and separate effect experiments which should be pursued systematically and with clearly defined objectives. It is desirable that some international program is developed in this area. This paper is concentrated on those items in the thermal-hydraulic area which eventually determine the quality of future accident analysis codes.

Views on the future of thermal hydraulic modeling

International Nuclear Information System (INIS)

Ishii, M.

1997-01-01

It is essential for the U.S. NRC to sustain the highest level of the thermal-hydraulics and reactor safety research expertise and continuously improve their accident analysis capability. Such expertise should span over four different areas which are strongly related to each other. These are: (1) Reactor Safety Code Development, (2) Two-phase Flow Modeling, (3) Instrumentation and Fundamental Experimental Research, and (4) Separate Effect and Integral Test. The NRC is already considering a new effort in the area of advanced thermal-hydraulics effort. Its success largely depends on the availability of a significantly improved two-phase flow formulation and constitutive relations supported by detailed experimental data. Therefore, it is recommended that the NRC start significant research efforts in the areas of two-phase flow modeling, instrumentation, basic and separate effect experiments which should be pursued systematically and with clearly defined objectives. It is desirable that some international program is developed in this area. This paper is concentrated on those items in the thermal-hydraulic area which eventually determine the quality of future accident analysis codes

Thermal performance and efficiency of supercritical nuclear reactors

International Nuclear Information System (INIS)

Romney Duffey; Tracy Zhou; Hussam Khartabil

2009-01-01

The paper reviews the major advances and innovative aspects of the thermal performance of recent concepts for super-critical water-cooled nuclear reactors (SCWR). The concepts are based on the extensive experience in the thermal power industry with super and ultra-supercritical boilers and turbines. The challenges and goals of increased efficiency, reduced cost, enhanced safety and co-generation have been pursued over the last ten years, and have resulted both in viable concepts and a vibrant defined R and D effort. The supercritical concept has wide acceptance among industry, as it reflects standard engineering practices and current thermal plant technology that is being already deployed. The SCWR concept represents a continuous development of water-cooled reactor technology, which utilizes the best and latest advances made in the thermal power industry. (author)

The detector safety system for LHC experiments

CERN Document Server

Schmeling, Sascha; Lüders, S; Morpurgo, Giulio

2004-01-01

The Detector Safety System (DSS), currently being developed at CERN under the auspices of the Joint Controls Project (JCOP), will be responsible for assuring the protection of equipment for the four Large Hadron Collider (LHC)**1 experiments. Thus, the DSS will require a high degree of both availability and reliability. After evaluation of various possible solutions, a prototype is being built based on a redundant Siemens PLC**2 front-end, to which the safety- critical part of the DSS task is delegated. This is then supervised by a PVSS**3 SCADA**4 system via an OPC**5 server. The PLC front-end is capable of running autonomously and of automatically taking predefined protective actions whenever required. The supervisory layer provides the operator with a status display and with limited online reconfiguration capabilities. Configuration of the code running in the PLCs will be completely data driven via the contents of a "configuration database." Thus, the DSS can easily adapt to the different and constantly ev...

Thermal hydraulic behavior of sub-assembly local blockage in China experiment fast reactor

International Nuclear Information System (INIS)

Yang Zhimin

2000-01-01

The geometrical parameter ratio of pitch to diameter of China Experiment Fast Reactor (CEFR) subassembly is 1,167. To address the thermal hydraulic behavior of subassembly local blockage which may be caused by deformation of cladding due to severe swelling and thermal stresses and by space swirl loosening etc., the porous numerical model and SIMPLE-P code used to solve Navier-Stokes and energy equations in porous medium was developed, and the bundle experiment with 19 pins with 24 subchannels blocked in the sodium coolant was carried on in China Institute of Atomic Energy (CIAE). The comparison of code predictions against experiments (including non-blockage and ten blockage conditions) seems well. The thermal hydraulic behavior of fuel subassembly with 61 fuel pins blockage of CEFR is calculated with SIMPLE-P code. The results indicate that the maximum temperature is 815 deg. C when the blockage area is about 37% (54 central subchannels are blocked). In this case the cladding won't be damaged and no sodium coolant boiling takes place. (author)

To improve nuclear plant safety by learning from accident's experience

International Nuclear Information System (INIS)

Matsumoto, Hidezo; Kida, Masanori; Kato, Hiroyuki; Hara, Shin-ichi

1994-01-01

The ultimate goal of this study is to produce an expert system that enables the experience (records and information) gained from accidents to be put to use towards improving nuclear plant safety. A number of examples have been investigated, both domestic and overseas, in which experience gained from accidents was utilized by utilities in managing and operating their nuclear power stations to improve safety. The result of investigation has been used to create a general 'basic flow' to make the best use of experience. The ultimate goal is achieved by carrying out this 'basic flow' with artificial intelligence (AI). To do this, it is necessary (1) to apply language analysis to process the source information (primary data base; domestic and overseas accident's reports) into the secondary data base, and (2) to establish an expert system for selecting (screening) significant events from the secondary data base. In the processing described in item (1), a multi-lingual thesaurus for nuclear-related terms become necessary because the source information (primary data bases) itself is multi-lingual. In the work described in item (2), the utilization of probabilistic safety assessment (PSA), for example, is a candidate method for judging the significance of events. Achieving the goal thus requires developing various new techniques. As the first step of the above long-term study project, this report proposes the 'basic flow' and presents the concept of how the nuclear-related AI can be used to carry out this 'basic flow'. (author)

TMX tandem-mirror experiments and thermal-barrier theoretical studies

International Nuclear Information System (INIS)

Simonen, T.C.; Baldwin, D.E.; Allen, S.L.

1982-01-01

This paper describes recent analysis of energy confinement in the Tandem Mirror Experiment (TMX). TMX data also indicates that warm plasma limits the amplitude of the anisotropy driven Alfven ion cyclotron (AIC) mode. Theoretical calculations show strong AIC stabilization with off-normal beam injection as planned in TMX-U and MFTF-B. This paper reports results of theoretical analysis of hot electrons in thermal barriers including electron heating calculations by Monte Carlo and Fokker-Planck codes and analysis of hot electron MHD and microinstability. Initial results from the TMX-U experiment are presented which show the presence of sloshing ions

Application of the Dragon reactor experiment to the safety evaluation of current HTR systems

International Nuclear Information System (INIS)

Ashworth, F.P.O.; Faircloth, R.L.

1976-01-01

An important component of the confidence required for the safety assessment of high-temperature reactors is the experimental proof of phenomena such as fission product release or core corrosion. The most convincing experiments are those carried out in a reactor. This paper outlines the scope of experiments relevant to safety which can be done in the Dragon Reactor Experiment and describes as an example the experimental campaign and the current outcome of the work on validating the predictions of caesium release and migration. (author)

Daily Thermal Predictions of the AGR-1 Experiment with Gas Gaps Varying with Time

Energy Technology Data Exchange (ETDEWEB)

Grant Hawkes; James Sterbentz; John Maki; Binh Pham

2012-06-01

A new daily as-run thermal analysis was performed at the Idaho National Laboratory on the Advanced Gas Reactor (AGR) test experiment number one at the Advanced Test Reactor (ATR). This thermal analysis incorporates gas gaps changing with time during the irradiation experiment. The purpose of this analysis was to calculate the daily average temperatures of each compact to compare with experimental results. Post irradiation examination (PIE) measurements of the graphite holder and fuel compacts showed the gas gaps varying from the beginning of life. The control temperature gas gap and the fuel compact – graphite holder gas gaps were linearly changed from the original fabrication dimensions, to the end of irradiation measurements. A steady-state thermal analysis was performed for each daily calculation. These new thermal predictions more closely match the experimental data taken during the experiment than previous analyses. Results are presented comparing normalized compact average temperatures to normalized log(R/B) Kr-85m. The R/B term is the measured release rate divided by the predicted birth rate for the isotope Kr-85m. Correlations between these two normalized values are presented.

A two-compartment thermal-hydraulic experiment (LACE-LA4) analyzed by ESCADRE code

International Nuclear Information System (INIS)

Passalacqua, R.

1994-01-01

Large scale experiments show that whenever a Loss of Coolant Accident (LOCA) occurs, water pools are generated. Stratifications of steam saturated gas develop above water pools causing a two-compartment thermal-hydraulics. The LACE (LWR Advanced Containment Experiment) LA4 experiment, performed at the Hanford Engineering Development Laboratory (HEDL), exhibited a strong stratification, at all times, above a growing water pool. JERICHO and AEROSOLS-B2 are part of the ESCADRE code system (Ensemble de Systemes de Codes d'Analyse d'accident Des Reacteurs A Eau), a tool for evaluating the response of a nuclear plant to severe accidents. These two codes are here used to simulate respectively the thermal-hydraulics and the associated aerosol behavior. Code results have shown that modelling large containment thermal-hydraulics without taking account of the stratification phenomenon leads to large overpredictions of containment pressure and temperature. If the stratification is modelled as a zone with a higher steam condensation rate and a higher thermal resistance, ESCADRE predictions match quite well experimental data. The stratification thermal-hydraulics is controlled by power (heat fluxes) repartition in the lower compartment between the water pool and the nearby walls. Therefore the total, direct heat exchange between the two compartment is reduced. Stratification modelling is believed to be important for its influence on aerosol behavior: aerosol deposition through the inter-face of the two subcompartments is improved by diffusiophoresis and thermophoresis. In addition the aerosol concentration gradient, through the stratification, will cause a driving force for motion of smaller particles towards the pool. (author)

The French nuclear safety authority's experience with radioactive transport inspection

International Nuclear Information System (INIS)

Jacob, E.; Aguilar, J.

2004-01-01

About 300,000 radioactive material packages are transported annually in France. Most consist of radioisotopes for medical, pharmaceutical or industrial use. On the other hand, the nuclear industry deals with the transport of fuel cycle materials (uranium, fuel assemblies, etc.) and waste from power plants, reprocessing plants and research centers. France is also a transit country for shipments such as spent fuel packages from Switzerland or Germany, which are bound for Sellafield in Great Britain. The French nuclear safety authority (DGSNR: Directorate General for Nuclear Safety and Radioprotection) has been responsible since 1997 for the safety of radioactive material transport. This paper presents DGNSR's experience with transport inspection: a feedback of key points based on 300 inspections achieved during the past five years is given

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

Recommended nuclear criticality safety experiments in support of the safe transportation of fissile material

International Nuclear Information System (INIS)

Tollefson, D.A.; Elliott, E.P.; Dyer, H.R.; Thompson, S.A.

1993-01-01

Validation of computer codes and nuclear data (cross-section) libraries using benchmark quality critical (or certain subcritical) experiments is an essential part of a nuclear criticality safety evaluation. The validation results establish the credibility of the calculational tools for use in evaluating a particular application. Validation of the calculational tools is addressed in several American National Standards Institute/American Nuclear Society (ANSI/ANS) standards, with ANSI/ANS-8.1 being the most relevant. Documentation of the validation is a required part of all safety analyses involving significant quantities of fissile materials. In the case of transportation of fissile materials, the safety analysis report for packaging (SARP) must contain a thorough discussion of benchmark experiments, detailing how the experiments relate to the significant packaging and contents materials (fissile, moderating, neutron absorbing) within the package. The experiments recommended in this paper are needed to address certain areas related to transportation of unirradiated fissile materials in drum-type containers (packagings) for which current data are inadequate or are lacking

Transient analysis for resolving safety issues

International Nuclear Information System (INIS)

Chao, J.; Layman, W.

1987-01-01

The Nuclear Safety Analysis Center (NSAC) has a Generic Safety Analysis Program to help resolve high priority generic safety issues. This paper describes several high priority safety issues considered at NSAC and how they were resolved by transient analysis using thermal hydraulics and neutronics codes. These issues are pressurized thermal shock (PTS), anticipated transients without scram (ATWS), steam generator tube rupture (SGTR), and reactivity transients in light of the Chernobyl accident

Thermal calculations pertaining to experiments in the Yucca Mountain Exploratory Shaft

International Nuclear Information System (INIS)

Montan, D.N.

1986-03-01

A series of thermal calculations have been presented that appear to satisfy the needs for design of the Yucca Mountain Exploratory Shaft Tests. The accuracy of the modeling and calculational techniques employed probably exceeds the accuracy of the thermal properties used. The rather close agreement between simple analytical methods (the PLUS Family) and much more complex methods (TRUMP) suggest that the PLUS Family might be appropriate during final design to model, in a single calculation, the entire test array and sequence. Before doing further calculations it is recommended that all available thermal property information be critically evaluated to determine ''best'' values to be used for conductivity and saturation. Another possibility is to design one or more of the test sequences to approximately duplicate the early phase of Heater Test 1. In that experiment an unplanned power outage for about two days that occurred a week into the experiment gave extremely useful data from which to determine the conductivity and diffusivity. In any case we urge that adequate, properly calibrated instrumentation with data output available on a quasi-real time basis be installed. This would allow us to take advantage of significant power changes (planned or not) and also help ''steer'' the tests to desired temperatures. Finally, it should be kept in mind that the calculations presented here are strictly thermal. No hydrothermal effects due to liquid and vapor pressures have been considered

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)

Planning the Unplanned Experiment: Towards Assessing the Efficacy of Standards for Safety-Critical Software

Science.gov (United States)

Graydon, Patrick J.; Holloway, C. M.

2015-01-01

Safe use of software in safety-critical applications requires well-founded means of determining whether software is fit for such use. While software in industries such as aviation has a good safety record, little is known about whether standards for software in safety-critical applications 'work' (or even what that means). It is often (implicitly) argued that software is fit for safety-critical use because it conforms to an appropriate standard. Without knowing whether a standard works, such reliance is an experiment; without carefully collecting assessment data, that experiment is unplanned. To help plan the experiment, we organized a workshop to develop practical ideas for assessing software safety standards. In this paper, we relate and elaborate on the workshop discussion, which revealed subtle but important study design considerations and practical barriers to collecting appropriate historical data and recruiting appropriate experimental subjects. We discuss assessing standards as written and as applied, several candidate definitions for what it means for a standard to 'work,' and key assessment strategies and study techniques and the pros and cons of each. Finally, we conclude with thoughts about the kinds of research that will be required and how academia, industry, and regulators might collaborate to overcome the noted barriers.

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

Safety

International Nuclear Information System (INIS)

2001-01-01

This annual report of the Senior Inspector for the Nuclear Safety, analyses the nuclear safety at EDF for the year 1999 and proposes twelve subjects of consideration to progress. Five technical documents are also provided and discussed concerning the nuclear power plants maintenance and safety (thermal fatigue, vibration fatigue, assisted control and instrumentation of the N4 bearing, 1300 MW reactors containment and time of life of power plants). (A.L.B.)

Screening of Gas-Cooled Reactor Thermal-Hydraulic and Safety Analysis Tools and Experimental Database

International Nuclear Information System (INIS)

Lee, Won Jae; Kim, Min Hwan; Lee, Seung Wook

2007-08-01

This report is a final report of I-NERI Project, 'Screening of Gas-cooled Reactor Thermal Hydraulic and Safety Analysis Tools and Experimental Database 'jointly carried out by KAERI, ANL and INL. In this study, we developed the basic technologies required to develop and validate the VHTR TH/safety analysis tools and evaluated the TH/safety database information. The research tasks consist of; 1) code qualification methodology (INL), 2) high-level PIRTs for major nucleus set of events (KAERI, ANL, INL), 3) initial scaling and scoping analysis (ANL, KAERI, INL), 4) filtering of TH/safety tools (KAERI, INL), 5) evaluation of TH/safety database information (KAERI, INL, ANL) and 6) key scoping analysis (KAERI). The code qualification methodology identifies the role of PIRTs in the R and D process and the bottom-up and top-down code validation methods. Since the design of VHTR is still evolving, we generated the high-level PIRTs referencing 600MWth block-type GT-MHR and 400MWth pebble-type PBMR. Nucleus set of events that represents the VHTR safety and operational transients consists of the enveloping scenarios of HPCC (high pressure conduction cooling: loss of primary flow), LPCC/Air-Ingress (low pressure conduction cooling: loss of coolant), LC (load changes: power maneuvering), ATWS (anticipated transients without scram: reactivity insertion), WS (water ingress: water-interfacing system break) and HU (hydrogen-side upset: loss of heat sink). The initial scaling analysis defines dimensionless parameters that need to be reflected in mixed convection modeling and the initial scoping analysis provided the reference system transients used in the PIRTs generation. For the PIRTs phenomena, we evaluated the modeling capability of the candidate TH/safety tools and derived a model improvement need. By surveying and evaluating the TH/safety database information, a tools V and V matrix has been developed. Through the key scoping analysis using available database, the modeling

Screening of Gas-Cooled Reactor Thermal-Hydraulic and Safety Analysis Tools and Experimental Database

Energy Technology Data Exchange (ETDEWEB)

Lee, Won Jae; Kim, Min Hwan; Lee, Seung Wook (and others)

2007-08-15

This report is a final report of I-NERI Project, 'Screening of Gas-cooled Reactor Thermal Hydraulic and Safety Analysis Tools and Experimental Database 'jointly carried out by KAERI, ANL and INL. In this study, we developed the basic technologies required to develop and validate the VHTR TH/safety analysis tools and evaluated the TH/safety database information. The research tasks consist of; 1) code qualification methodology (INL), 2) high-level PIRTs for major nucleus set of events (KAERI, ANL, INL), 3) initial scaling and scoping analysis (ANL, KAERI, INL), 4) filtering of TH/safety tools (KAERI, INL), 5) evaluation of TH/safety database information (KAERI, INL, ANL) and 6) key scoping analysis (KAERI). The code qualification methodology identifies the role of PIRTs in the R and D process and the bottom-up and top-down code validation methods. Since the design of VHTR is still evolving, we generated the high-level PIRTs referencing 600MWth block-type GT-MHR and 400MWth pebble-type PBMR. Nucleus set of events that represents the VHTR safety and operational transients consists of the enveloping scenarios of HPCC (high pressure conduction cooling: loss of primary flow), LPCC/Air-Ingress (low pressure conduction cooling: loss of coolant), LC (load changes: power maneuvering), ATWS (anticipated transients without scram: reactivity insertion), WS (water ingress: water-interfacing system break) and HU (hydrogen-side upset: loss of heat sink). The initial scaling analysis defines dimensionless parameters that need to be reflected in mixed convection modeling and the initial scoping analysis provided the reference system transients used in the PIRTs generation. For the PIRTs phenomena, we evaluated the modeling capability of the candidate TH/safety tools and derived a model improvement need. By surveying and evaluating the TH/safety database information, a tools V and V matrix has been developed. Through the key scoping analysis using available database, the

Experience of Tecnatom in Developing a Strong Leadership for Safety and Performance

International Nuclear Information System (INIS)

González, F.; Villadóniga, J. I.

2016-01-01

This paper presents experience and insights of Tecnatom in the support of internal and external clients to develop a strong Leadership for Safety. Several cases are presented briefly: (a) The leadership and culture change activities for a utility, a radwaste company, and for Tecnatom itself. One important characteristic of the work performed is the detailed consideration of the underlying organizational culture that underpins the safety culture. Measurable improvements have been achieved and some of the key insights are shared in this paper. (b) The development and implementation of a leadership model with 17 competencies, including safety explicitly. One benefit of this model is that allows to perform a quantitative assessment of leadership effectiveness, something vital to be able to ensure that leadership development actions are truly supporting safety. The model uses an approach to development oriented to strengths and the use of companion competencies to further develop leadership. Moreover it aims to produce significant improvements on safety but also on performance, since both are not competing goals when the proper leadership model is selected. The training material prepared was shortlisted in the 2014 Nuclear Training Awards. (c) The design and implementation of a training development program on Safety Culture, and required competencies of Leadership, for Top Managers of the nuclear industry, as part of the project NUSHARE of the European Commission’s 7th research framework program. The program is sensible to the reduced time availability of Top Managers and uses a combination of learning approaches (webinars, micro-elearnings, web meetings) that provide higher flexibility for the learner, but complemented with other proven methods (group dialog, journaling, mentoring, etc.) to ensure that the program is effective. All these experiences reveal that to improve the organizational Safety Culture we need to enhance Leadership for Safety and Performance

Patients' experiences of safety during haemodialysis treatment--a qualitative study

NARCIS (Netherlands)

Lovink, Marleen H; Kars, Marijke C; de Man-van Ginkel, Janneke M; Schoonhoven, Lisette

2015-01-01

AIM: To explore the experiences of safety of adult patients during their haemodialysis treatment. BACKGROUND: Haemodialysis is a complex treatment with a risk for harm that causes anxiety among many patients. To date, no in-depth study of haemodialysis patients' emotional responses to conditions of

Patients' experiences of safety during haemodialysis treatment - a qualitative study

NARCIS (Netherlands)

Lovink, M.H.; Kars, M.C.; Man-van Ginkel, J.M. de; Schoonhoven, L.

2015-01-01

AIM: To explore the experiences of safety of adult patients during their haemodialysis treatment. BACKGROUND: Haemodialysis is a complex treatment with a risk for harm that causes anxiety among many patients. To date, no in-depth study of haemodialysis patients' emotional responses to conditions of

Safety research programs sponsored by Office of Nuclear Regulatory Research. Volume 3. No. 2. Quarterly progress report, April 1-June 30, 1983

Energy Technology Data Exchange (ETDEWEB)

Bari, R A; Cerbone, R J; Ginsberg, T; Greene, G A; Guppy, J G; Hall, R E; Luckas, Jr, W J; Reich, M; Saha, P; Sastre, C

1983-09-01

The projects reported are the following: HTGR Safety Evaluation, SSC development, Validation and Application, CRBR balance of plant modeling, thermal-hydraulic reactor safety experiments, LWR plant analyzer development, LWR code assessment and application, thermal reactor code development (RAMONA-3B); stress corrosion cracking of PWR steam generator tubing, bolting failure analysis, probability based load combinations for design of category I structures, mechanical piping benchmark problems; human error data for nuclear power plant safety related events, criteria for human engineering regulatory guides and human factors in nuclear power plant safeguards.

Thermal comfort and safety of cotton blankets warmed at 130°F and 200°F.

Science.gov (United States)

Kelly, Patricia A; Cooper, Susan K; Krogh, Mary L; Morse, Elizabeth C; Crandall, Craig G; Winslow, Elizabeth H; Balluck, Julie P

2013-12-01

In 2009, the ECRI Institute recommended warming cotton blankets in cabinets set at 130°F or less. However, there is limited research to support the use of this cabinet temperature. To measure skin temperatures and thermal comfort in healthy volunteers before and after application of blankets warmed in cabinets set at 130 and 200°F, respectively, and to determine the time-dependent cooling of cotton blankets after removal from warming cabinets set at the two temperatures. Prospective, comparative, descriptive. Participants (n = 20) received one or two blankets warmed in 130 or 200°F cabinets. First, skin temperatures were measured, and thermal comfort reports were obtained at fixed timed intervals. Second, blanket temperatures (n = 10) were measured at fixed intervals after removal from the cabinets. No skin temperatures approached levels reported in the literature that cause epidermal damage. Thermal comfort reports supported using blankets from the 200°F cabinet, and blankets lost heat quickly over time. We recommend warming cotton blankets in cabinets set at 200°F or less to improve thermal comfort without compromising patient safety. Copyright © 2013 American Society of PeriAnesthesia Nurses. Published by Elsevier Inc. All rights reserved.

Pressure locking and thermal binding of gate valves

Energy Technology Data Exchange (ETDEWEB)

Kelly, E.M.

1996-12-01

Pressure locking and thermal binding represent potential common mode failure mechanisms that can cause safety-related power-operated gate valves to fail in the closed position, thus rendering redundant safety-related systems incapable of performing their safety functions. Supplement 6 to Generic Letter 89-10, {open_quotes}Safety-Related Motor-Operated Gate Valve Testing and Surveillance,{close_quotes} provided an acceptable approach to addressing pressure locking and thermal binding of gate valves. More recently, the NRC has issued Generic Letter 95-07, {open_quotes}Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves,{close_quotes} to request that licensees take certain actions to ensure that safety-related power-operated gate valves that are susceptible to pressure locking or thermal binding are capable of performing their safety functions within the current licensing bases. Over the past two years, several plants in Region I determined that valves in certain systems were potentially susceptible to pressure locking and thermal binding, and have taken various corrective actions. The NRC Region I Systems Engineering Branch has been actively involved in the inspection of licensee actions in response to the pressure locking and thermal binding issue. Region I continues to maintain an active involvement in this area, including participation with the Office of Nuclear Reactor Regulation in reviewing licensee responses to Generic Letter 95-07.

JET-ISX-B beryllium limiter experiment safety analysis report and operational safety requirements

International Nuclear Information System (INIS)

Edmonds, P.H.

1985-09-01

An experiment to evaluate the suitability of beryllium as a limiter material has been completed on the ISX-B tokamak. The experiment consisted of two phases: (1) the initial operation and characterization in the ISX experiment, and a period of continued operation to the specified surface fluence (10 22 atoms/cm 2 ) of hydrogen ions; and (2) the disassembly, decontamination, or disposal of the ISX facility. During these two phases of the project, the possibility existed for beryllium and/or beryllium oxide powder to be produced inside the vacuum vessel. Beryllium dust is a highly toxic material, and extensive precautions are required to prevent the release of the beryllium into the experimental work area and to prevent the contamination of personnel working on the device. Details of the health hazards associated with beryllium and the appropriate precautions are presented. Also described in appendixes to this report are the various operational safety requirements for the project

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

International Nuclear Information System (INIS)

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

1986-05-01

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

Crack propagation behavior of TiN coatings by laser thermal shock experiments

International Nuclear Information System (INIS)

Choi, Youngkue; Jeon, Seol; Jeon, Min-seok; Shin, Hyun-Gyoo; Chun, Ho Hwan; Lee, Youn-seoung; Lee, Heesoo

2012-01-01

Highlights: ► The crack propagation behavior of TiN coating after laser thermal shock experiment was observed by using FIB and TEM. ► Intercolumnar cracks between TiN columnar grains were predominant cracking mode after laser thermal shock. ► Cracks were propagated from the coating surface to the substrate at low laser pulse energy and cracks were originated at coating-substrate interface at high laser pulse energy. ► The cracks from the interface spread out transversely through the weak region of the columnar grains by repetitive laser shock. - Abstract: The crack propagation behavior of TiN coatings, deposited onto 304 stainless steel substrates by arc ion plating technique, related to a laser thermal shock experiment has been investigated using focused ion beam (FIB) and transmission electron microscopy (TEM). The ablated regions of TiN coatings by laser ablation system have been investigated under various conditions of pulse energies and number of laser pulses. The intercolumnar cracks were predominant cracking mode following laser thermal shock tests and the cracks initiated at coating surface and propagated in a direction perpendicular to the substrate under low loads conditions. Over and above those cracks, the cracks originated from coating-substrate interface began to appear with increasing laser pulse energy. The cracks from the interface also spread out transversely through the weak region of the columnar grains by repetitive laser shock.

Proceedings of the workshop on integral experiment covariance data for critical safety validation

Energy Technology Data Exchange (ETDEWEB)

Stuke, Maik (ed.)

2016-04-15

For some time, attempts to quantify the statistical dependencies of critical experiments and to account for them properly in validation procedures were discussed in the literature by various groups. Besides the development of suitable methods especially the quality and modeling issues of the freely available experimental data are in the focus of current discussions, carried out for example in the Expert Group on Uncertainty Analysis for Criticality Safety Assessment (UACSA) of the OECD-NEA Nuclear Science Committee. The same committee compiles and publishes also the freely available experimental data in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. Most of these experiments were performed as series and might share parts of experimental setups leading to correlated results. The quality of the determination of these correlations and the underlying covariance data depend strongly on the quality of the documentation of experiments.

Proceedings of the workshop on integral experiment covariance data for critical safety validation

International Nuclear Information System (INIS)

Stuke, Maik

2016-04-01

For some time, attempts to quantify the statistical dependencies of critical experiments and to account for them properly in validation procedures were discussed in the literature by various groups. Besides the development of suitable methods especially the quality and modeling issues of the freely available experimental data are in the focus of current discussions, carried out for example in the Expert Group on Uncertainty Analysis for Criticality Safety Assessment (UACSA) of the OECD-NEA Nuclear Science Committee. The same committee compiles and publishes also the freely available experimental data in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. Most of these experiments were performed as series and might share parts of experimental setups leading to correlated results. The quality of the determination of these correlations and the underlying covariance data depend strongly on the quality of the documentation of experiments.

Numerical analysis of two experiments related to thermal fatigue

Energy Technology Data Exchange (ETDEWEB)

Bieder, Ulrich; Errante, Paolo [DEN-STMF, Commissariat a l' Energie Atomique et aux Energies Alternatives, Universite Paris-Saclay, Gif-sur-Yvette (France)

2017-06-15

Jets in cross flow are of fundamental industrial importance and play an important role in validating turbulence models. Two jet configurations related to thermal fatigue phenomena are investigated: • T-junction of circular tubes where a heated jet discharges into a cold main flow and • Rectangular jet marked by a scalar discharging into a main flow in a rectangular channel. The T-junction configuration is a classical test case for thermal fatigue phenomena. The Vattenfall T-junction experiment was already subject of an OECD/NEA benchmark. A LES modelling and calculation strategy is developed and validated on this data. The rectangular-jet configuration is important for basic physical understanding and modelling and has been analyzed experimentally at CEA. The experimental work was focused on turbulent mixing between a slightly heated rectangular jet which is injected perpendicularly into the cold main flow of a rectangular channel. These experiments are analyzed for the first time with LES. The overall results show a good agreement between the experimental data and the CFD calculation. Mean values of velocity and temperature are well captured by both RANS calculation and LES. The range of critical frequencies and their amplitudes, however, are only captured by LES.

Experiment to evaluate software safety

International Nuclear Information System (INIS)

Soubies, B.; Henry, J.Y.

1994-01-01

The process of licensing nuclear power plants for operation consists of mandatory steps featuring detailed examination of the instrumentation and control system by the safety authorities, including softwares. The criticality of these softwares obliges the manufacturer to develop in accordance with the IEC 880 standard 'Computer software in nuclear power plant safety systems' issued by the International Electronic Commission. The evaluation approach, a two-stage assessment is described in detail. In this context, the IPSN (Institute of Protection and Nuclear Safety), the technical support body of the safety authority uses the MALPAS tool to analyse the quality of the programs. (R.P.). 4 refs

Experiences in assessing safety culture

International Nuclear Information System (INIS)

Spitalnik, J.

2002-01-01

Based on several Safety Culture self-assessment applications in nuclear organisations, the paper stresses relevant aspects to be considered when programming an assessment of this type. Reasons for assessing Safety Culture, basic principles to take into account, necessary resources, the importance of proper statistical analyses, the feed-back of results, and the setting up of action plans to enhance Safety Culture are discussed. (author)

Methodical Specifics of Thermal Experiments with Thin Carbon Reinforced Plates

Directory of Open Access Journals (Sweden)

O. V. Denisov

2015-01-01

Full Text Available Polymer composite materials (CM are widely used in creation of large space constructions, especially reflectors of space antennas. Composite materials should provide high level of specific stiffness and strength for space structures. Thermal conductivity in reinforcement plane is a significant factor in case of irregular heating space antennas. Nowadays, data on CM reinforcement plane thermal conductivity are limited and existing methods of its defining are imperfect. Basically, traditional methods allow us to define thermal conductivity in perpendicular direction towards the reinforcement plane on the samples of round or rectangular plate. In addition, the thickness of standard samples is larger than space antenna thickness. Consequently, new methods are required. Method of contact heating, which was developed by BMSTU specialists with long hollow carbon beam, could be a perspective way. This article is devoted to the experimental method of contact heating on the thin carbon plates.Thermal tests were supposed to provide a non-stationary temperature field with a gradient being co-directional with the plane reinforcement in the material sample. Experiments were conducted in vacuum chamber to prevent unstructured convection. Experimental thermo-grams processing were calculated by 1-d thermal model for a thin plate. Influence of uncertainty of experimental parameters, such as (radiation emission coefficients of sample surface, glue, temperature sensors and uncertainty of sensors placement on the result of defined thermal conductivity has been estimated. New data on the thermal conductivity in reinforcement plane were obtained within 295 - 375 K temperature range, which can be used to design and develop reflectors of precision space antennas. In the future it is expedient to conduct tests of thin-wall plates from carbon fiber-reinforced plastic in wide temperature range, especially in the low-range temperatures.

Current MitraClip experience, safety and feasibility in the Netherlands

NARCIS (Netherlands)

Rahhab, Z.; Kortlandt, F. A.; Velu, J. F.; Schurer, R. A. J.; Delgado, V.; Tonino, P.; Boven, A. J.; van den Branden, B. J. L.; Kraaijeveld, A. O.; Voskuil, M.; Hoorntje, J.; van Wely, M. [=Marleen; van Houwelingen, K.; Bleeker, G. B.; Rensing, B.; Kardys, I.; Baan, J.; van der Heyden, J. A. S.; van Mieghem, N. M.

2017-01-01

Purpose Data on MitraClip procedural safety and efficacy in the Netherlands are scarce. We aim to provide an overview of the Dutch MitraClip experience. Methods We pooled anonymised demographic and procedural data of 1151 consecutive MitraClip patients, from 13 Dutch hospitals. Data was collected by

CFX-10 Analysis of the High Temperature Thermal- Chemical Experiment (CS28-2)

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyoung Tae; Park, Joo Hwan; Rhee, Bo Wook

2008-02-15

A Computational Fluid Dynamics (CFD) model of a post-blowdown fuel channel analysis for aged CANDU reactors with crept pressure tube has been developed, and validated against a high temperature thermal-chemical experiment: CS28-2. The CS28-2 experiment is one of three series of experiments to simulate the thermal-chemical behavior of a 28-element fuel channel at a high temperature and a low steam flow rate which may occur in severe accident conditions such as a LBLOCA (Large Break Loss of Coolant Accident) of CANDU reactors. Pursuant to the objective of this study, the current study has focused on understanding the involved phenomena such as the thermal radiation and convection heat transfer, and the high temperature zirconium-steam reaction in a multi-ring geometry. Therefore, a zirconium-steam oxidation model based on a parabolic rate law was implemented into the CFX-10 code, which is a commercial CFD code offered from ANSYS Inc., and other heat transfer mechanisms in the 28-element fuel channel were modeled by the original CFX-10 heat transfer packages. To assess the capability of the CFX-10 code to model the thermal-chemical behavior of the 28-element fuel channel, the measured temperatures of the Fuel Element Simulators (FES) of three fuel rings in the test bundle and the pressure tube, and the hydrogen production in the CS28-2 experiment were compared with the CFX-10 predictions. In spite of some discrepancy between the measurement data and CFX results, it was found that the CFX-10 prediction based on the Urbanic-Heidrick correlation of the zirconium-steam reaction as well as the Discrete Transfer Model for a radiation heat transfer among the FES of three rings and the pressure tube are quite accurate and sound even for the offset a cluster fuel bundle of an aged fuel channel.

CFX-10 Analysis of the High Temperature Thermal- Chemical Experiment (CS28-2)

International Nuclear Information System (INIS)

Kim, Hyoung Tae; Park, Joo Hwan; Rhee, Bo Wook

2008-02-01

A Computational Fluid Dynamics (CFD) model of a post-blowdown fuel channel analysis for aged CANDU reactors with crept pressure tube has been developed, and validated against a high temperature thermal-chemical experiment: CS28-2. The CS28-2 experiment is one of three series of experiments to simulate the thermal-chemical behavior of a 28-element fuel channel at a high temperature and a low steam flow rate which may occur in severe accident conditions such as a LBLOCA (Large Break Loss of Coolant Accident) of CANDU reactors. Pursuant to the objective of this study, the current study has focused on understanding the involved phenomena such as the thermal radiation and convection heat transfer, and the high temperature zirconium-steam reaction in a multi-ring geometry. Therefore, a zirconium-steam oxidation model based on a parabolic rate law was implemented into the CFX-10 code, which is a commercial CFD code offered from ANSYS Inc., and other heat transfer mechanisms in the 28-element fuel channel were modeled by the original CFX-10 heat transfer packages. To assess the capability of the CFX-10 code to model the thermal-chemical behavior of the 28-element fuel channel, the measured temperatures of the Fuel Element Simulators (FES) of three fuel rings in the test bundle and the pressure tube, and the hydrogen production in the CS28-2 experiment were compared with the CFX-10 predictions. In spite of some discrepancy between the measurement data and CFX results, it was found that the CFX-10 prediction based on the Urbanic-Heidrick correlation of the zirconium-steam reaction as well as the Discrete Transfer Model for a radiation heat transfer among the FES of three rings and the pressure tube are quite accurate and sound even for the offset a cluster fuel bundle of an aged fuel channel

The development of safety cases for healthcare services: Practical experiences, opportunities and challenges

International Nuclear Information System (INIS)

Sujan, Mark; Spurgeon, Peter; Cooke, Matthew; Weale, Andy; Debenham, Philip; Cross, Steve

2015-01-01

There has been growing interest in the concept of safety cases for medical devices and health information technology, but questions remain about how safety cases can be developed and used meaningfully in the safety management of healthcare services and processes. The paper presents two examples of the development and use of safety cases at a service level in healthcare. These first practical experiences at the service level suggest that safety cases might be a useful tool to support service improvement and communication of safety in healthcare. The paper argues that safety cases might be helpful in supporting healthcare organisations with the adoption of proactive and rigorous safety management practices. However, it is also important to consider the different level of maturity of safety management and regulatory oversight in healthcare. Adaptations to the purpose and use of safety cases might be required, complemented by the provision of education to both practitioners and regulators. - Highlights: • Empirical description of safety case development at service level in healthcare. • Safety cases can support adoption of proactive and rigorous safety management. • Adaptation to purpose and use of safety cases might be required in healthcare. • Education should be provided to practitioners and regulators

French-Finnish colloquium on safety of French and Russian type nuclear power plants

International Nuclear Information System (INIS)

Lukka, M.; Jaervinen, M.; Minkkinen, P.; Ukkola, A.; Levomaeki, L.

1994-01-01

The French-Finnish Colloquium on Safety of French and Russian Type Nuclear Power Plants was held in June, 14th - 16th, 1994, in Lappeenranta, Finland. The main topics of the colloquium were: VVER and RBMK reactors; Industrial safety studies for VVER's in FRAMATOME; Structural safety analysis of Ignalina NPP; Thermalhydraulic system (BETHSY) and analytical experiments for French NPP; Test facilities simulating VVER plants during accidents; PACTEL - facility for VVER thermal hydraulics; High burn-up fuel and reactivity accidents; Overview of severe accident research at Nuclear Protection and Safety Institute of CEA; Research of severe accidents in Finland; Review of main activities concerning computer codes used for VVER thermal-hydraulic safety analysis in OKB Gidropress; CATHARE code; APROS computer code, new developments; TRIO and TOLBIAC computer codes; ESTET and N3S softwares; HEXTRAN - 3D reactor dynamics code for VVER accident analysis; An overview the boron dilution issue in PWRs; Boron mixing transients in a 900 MW PWR vessel for a reactor start-up operation; and Problem of boric acid dilution in IVO

Non-LTE considerations in spectral diagnostics of thermal transport and implosion experiments

International Nuclear Information System (INIS)

Epstein, R.; Skupsky, S.; Delettrez, J.; Yaakobi, B.

1984-01-01

Recent thermal-transport and target-implosion experiments have used the emission of radiation from highly-ionized ions to signal the advance of laser-driven heat fronts and to mark the trajectories and stagnation points of imploding shells. We examine the results of such experiments with particular attention given to non-LTE effects of non-Maxwellian electrons and of finite ionization times on the populations of signature-emitting atomic species and on the formation of signature spectra and x-ray images in these experiments

AST-500 safety analysis experience

Energy Technology Data Exchange (ETDEWEB)

Falikov, A A; Bakhmetiev, A M; Kuul, V S; Samoilov, O B [OKBM, Nizhny Novgorod (Russian Federation)

1997-09-01

Characteristic AST-type NHR safety features and requirements are described briefly. The main approaches and results of design and beyond-design accidents analyses for the AST-500 NHR, and the results of probabilistic safety assessments are considered. It is concluded that the AST-500 possesses a high safety level in virtue of the development and realization in the design of self-protection, passivity and defence-in-depth principles. (author). 9 refs, 2 figs.

Thermal testing of solid neutron shielding materials

International Nuclear Information System (INIS)

Boonstra, R.H.

1993-01-01

In May-June 1989 the first series of full-scale thermal tests was performed on three shielding materials: Bisco Products NS-4-FR, and Reactor Experiments RX-201 and RX-207. The tests are described in Thermal Testing of Solid Neutron Shielding Materials, GA-A19897, R.H. Boonstra, General Atomics (1990), and demonstrated the acceptability of these materials in a thermal accident. Subsequent design changes to the cask rendered these materials unattractive in terms of weight or adequate service temperature margin. For the second test series a material specification was developed for a polypropylene based neutron shield with a softening point of at least 280degF. Table 1 lists the neutron shield materials tested. The Envirotech and Bisco materials are not polypropylene, but were tested as potential backup materials in the event that a satisfactory polypropylene could not be found. The Bisco modified NS-4 and Reactor Experiments HMPP are both acceptable materials from a thermal accident standpoint for use in the shipping cask. Tests of the Kobe PP-R01 and Envirotech HDPE were stopped for safety reasons, due to inability to deal with the heavy smoke, before completion of the 30-minute heating phase. However these materials may prove satisfactory if they could undergo the complete heating. (J.P.N.)

Safety and operating experience at EBR-II: lessons for the future

International Nuclear Information System (INIS)

Sackett, J.I.; Golden, G.H.

1981-01-01

EBR-II is a small LMFBR power plant that has performed safely and reliably for 16 years. Much has been learned from operating it to facilitate the design, licensing, and operation of large commercial LMFBR power plants in the US. EBR-II has been found relatively easy to keep in conformity with evolving safety requirements, largely because of inherent safety features of the plant. Such features reduce dependence on active safety systems to protect against accidents. EBR-II has experienced a number of plant-transient incidents, some planned, others inadvertent; none has resulted in any significant plant damage. The operating experience with EBR-II has led to the formulation of an Operational Reliability Test Program (ORTP), aimed at showing inherently safe performance of fuel and plant systems

Applying the results of probablistic safety analysis of nuclear power plants: a survey of experience

International Nuclear Information System (INIS)

Andrews, W.B.; Herttrich, M.; Koeberlein, K.; Schwager.

1985-01-01

To date, discussions of the many different types of potential applications of PRA/PSA results and insights to safety-decision-making have been mainly theoretical. Various safety goals have been proposed as decision criteria. However, the discussion on the role of PRA/PSA and Safety Goals in safety-decision-making, especially in licensing, is controversial. A Working Group of the OECD Nuclear Energy Agency is completing a compilation and evaluation of real examples of past and present practical experience with the application of probabilistic methods in reactor safety decision-making, with the idea of developing a common understanding in this area. More than fifty different cases where PRA has influenced decision-making have been surveyed. These include, for example, regulatory changes, fixing of licensing requirements, plant specific modifications of design of operation, prioritization of safety issues and emergency planning. This feedback of experience - both positive and negative - with PRA/PSA applications is considered to provide guidance on how probabilistic approaches can be introduced into current safety practices, and on desirable future developments in probabilistic methods and specific PSA/PRA studies. Generic insights from the survey are given

Statistical safety evaluation of BWR turbine trip scenario using coupled neutron kinetics and thermal hydraulics analysis code SKETCH-INS/TRACE5.0

International Nuclear Information System (INIS)

Ichikawa, Ryoko; Masuhara, Yasuhiro; Kasahara, Fumio

2012-01-01

The Best Estimate Plus Uncertainty (BEPU) method has been prepared for the regulatory cross-check analysis at Japan Nuclear Energy Safety Organization (JNES) on base of the three-dimensional neutron-kinetics/thermal-hydraulics coupled code SKETCH-INS/TRACE5.0. In the preparation, TRACE5.0 is verified against the large-scale thermal-hydraulic tests carried out with NUPEC facility. These tests were focused on the pressure drop of steam-liquid two phase flow and void fraction distribution. From the comparison of the experimental data with other codes (RELAP5/MOD3.3 and TRAC-BF1), TRACE5.0 was judged better than other codes. It was confirmed that TRACE5.0 has high reliability for thermal hydraulics behavior and are used as a best-estimate code for the statistical safety evaluation. Next, the coupled code SKETCH-INS/TRACE5.0 was applied to turbine trip tests performed at the Peach Bottom-2 BWR4 Plant. The turbine trip event shows the rapid power peak due to the voids collapse with the pressure increase. The analyzed peak value of core power is better simulated than the previous version SKETCH-INS/TRAC-BF1. And the statistical safety evaluation using SKETCH-INS/TRACE5.0 was applied to the loss of load transient for examining the influence of the choice of sampling method. (author)

DASHBOARDS AND CONTROL CHARTS. EXPERIENCES IN IMPROVING SAFETY AT HANFORD WASHINGTON

International Nuclear Information System (INIS)

PREVETTE, S.S.

2006-01-01

The aim of this paper is to demonstrate the integration of safety methodology, quality tools, leadership, and teamwork at Hanford and their significant positive impact on safe performance of work. Dashboards, Leading Indicators, Control charts, Pareto Charts, Dr. W. Edward Deming's Red Bead Experiment, and Dr. Deming's System of Profound Knowledge have been the principal tools and theory of an integrated management system. Coupled with involved leadership and teamwork, they have led to significant improvements in worker safety and protection, and environmental restoration at one of the nation's largest nuclear cleanup sites

Review of the Thermal-Chemical Experiments for CANDU Fuel Channel

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyoung Tae; Min, Byung Joo; Park, Joo Hwan; Yoon, Churl; Rhee, Bo Wook

2005-08-15

In the present study, thermal-chemical experiments for CANDU channel analysis are reviewed. First, 11 experiments are identified from the present references and classified according to the number of heater rods, channel orientation, and degree of FES (Fuel Element Simulator) temperature rise during transient. The main configuration of the test rigs and the position of the measurement systems are identified. The experiments were generally conducted in three stages, a low-power, a high-power and a no-power stage. These test procedures are classified and described in this document. The experimental conditions for steam, coolant, and heat power are identified. The thermal properties of solid materials and fluids in the test apparatus are listed in the tables. From the review of the main test results, the following conclusions are to be obtained. Some of the reviewed experiments were not in the quasy-steady state conditions at a low-power stage and followed by a high-power stage. Zircaloy/steam reaction started when FES temperature were 800 .deg. C and escalated when temperature exceeded 1150 .deg. C. Uncontrolled temperature escalations due to Zircaloy/steam reaction were not observed when the FES temperature reached peak point (just below the melting point) and electric power to the test section shut off (self-sustaining Zircaloy/steam reaction). There were negligible circumferential temperature gradients in the FES bundle and pressure tube for the experiments performed in a vertical channel orientation. There were, however, noticeable circumferential gradients when the pressure tube was horizontal. These gradients were attributed to slumping of the FES bundle (sagging). Sagging of the bundle may have masked any buoyancy induced temperature gradients. Furthermore, the hot FES sagged towards the pressure tube transferring more heat to the pressure tube and increasing the temperature of the pressure tube.

Review of the Thermal-Chemical Experiments for CANDU Fuel Channel

International Nuclear Information System (INIS)

Kim, Hyoung Tae; Min, Byung Joo; Park, Joo Hwan; Yoon, Churl; Rhee, Bo Wook

2005-08-01

In the present study, thermal-chemical experiments for CANDU channel analysis are reviewed. First, 11 experiments are identified from the present references and classified according to the number of heater rods, channel orientation, and degree of FES (Fuel Element Simulator) temperature rise during transient. The main configuration of the test rigs and the position of the measurement systems are identified. The experiments were generally conducted in three stages, a low-power, a high-power and a no-power stage. These test procedures are classified and described in this document. The experimental conditions for steam, coolant, and heat power are identified. The thermal properties of solid materials and fluids in the test apparatus are listed in the tables. From the review of the main test results, the following conclusions are to be obtained. Some of the reviewed experiments were not in the quasy-steady state conditions at a low-power stage and followed by a high-power stage. Zircaloy/steam reaction started when FES temperature were 800 .deg. C and escalated when temperature exceeded 1150 .deg. C. Uncontrolled temperature escalations due to Zircaloy/steam reaction were not observed when the FES temperature reached peak point (just below the melting point) and electric power to the test section shut off (self-sustaining Zircaloy/steam reaction). There were negligible circumferential temperature gradients in the FES bundle and pressure tube for the experiments performed in a vertical channel orientation. There were, however, noticeable circumferential gradients when the pressure tube was horizontal. These gradients were attributed to slumping of the FES bundle (sagging). Sagging of the bundle may have masked any buoyancy induced temperature gradients. Furthermore, the hot FES sagged towards the pressure tube transferring more heat to the pressure tube and increasing the temperature of the pressure tube

The role of quality management in safety case development - Nagra's experience

International Nuclear Information System (INIS)

Schneider, Juerg W.; Zuidema, Piet

2014-01-01

This paper discusses the role of quality management (QM) in safety case development based on Nagra's experience from a broad range of projects. These include Project Gewahr (L/ILW and HLW, Nagra, 1985), the Wellenberg Project (L/ILW, Nagra, 1994), Project Opalinus Clay (HLW, Nagra, 2002a, 2002b), and recent project work needed in the context of the Swiss site selection process (L/ILW and HLW, Nagra, 2008a, 2008b, 2008c, 2010). Broadly speaking, Nagra's Quality Management policy is focused on ensuring: i) the quality of the disposal system (siting, design and implementation); ii) the quality of the underlying scientific understanding, which are seen as key elements of a credible safety case, along with the quality of the safety calculations themselves and of compiling the safety case, including the drawing of conclusions (Nagra, 2002a). All aspects of QM discussed in this paper should be seen in this context. (authors)

Thermal-hydraulic analysis code development and application to passive safety reactor at JAERI

International Nuclear Information System (INIS)

Araya, F.

1995-01-01

After a brief overview of safety assessment process, the author describes the LOCA analysis code system developed in JAERI. It comprises audit calculation code (WREM, WREM-J2, Japanese own code and BE codes (2D/3D, ICAP, ROSA). The codes are applied to development of Japanese passive safety reactor concept JPSR. Special attention is paid to the passive heat removal system and phenomena considered to occur under loss of heat sink event. Examples of LOCA analysis based on operation of JPSR for the cases of heat removal by upper RHR and heat removal from core to atmosphere are given. Experiments for multi-dimensional flow field in RPV and steam condensation in water pool are used for understanding the phenomena in passive safety reactors. The report is in a poster form only. 1 tab., 13 figs

Development of thermal-hydraulic models for the safety evaluation of CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Lee, Jae Young; Hwang, Gi Suk; Jung, Yun Sik [Handong Univ., Pohang (Korea, Republic of); No, Hee Cheon; Moon, Young Min [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

2003-03-15

The objective of the present research is to evaluate the safety analysis for CANDU and to improve the Horizontal Stratification Entrainment Model (HSEM) of RELAP5/MOD3.3. This report includes two items : the one is the development of experimental facility for the safety evaluation of CANDU, the other is the results of comparison with the existing correlations and data. The literature reviews are performed and the database for previous off-take experiments are built. By a survey of state-of-the-articles, the deficiencies of previous works and limitations of existing models are examined. The hydraulic behavior branching through the feeder pipes from the header pipe is analyzed and the test facility of off-take experiment is designed and manufactured as the prototype CANDU6, by a proper scaling methodologies. The test facility contains various branch pipes not only for three directions (top, side and bottom), but for arbitrary directions. The experiments about the onset of entrainment and branch quality only for three directions (top, side and bottom) are carried out by using air-water as working fluids. On the whole, the existing correlations predict the present experimental results well branch quality, entrainment, which validates the availability of experimental facility and methodology. Especially, for the branch quality with top and bottom branches, the different results are shown because of the unstable flow regimes in the horizontal pipe and the different branch diameters. The deficiencies of previous works and limitations of existing models are considered. The off-take experiment for arbitrary branch angles continues as the next year research.

Summary of LWR safety research in the USA

International Nuclear Information System (INIS)

Murley, T.E.; Tong, L.S.; Bennett, G.L.

1977-01-01

The U.S. Nuclear Regulatory Commission's water reactor safety research program is described and the basic results are presented. The USNRC water reactor safety research program consists of five basic research areas: integrity of vessel and piping, thermal-hydraulic test, fuel rod behaviour, code development and verification, and reactor operational safety. Results from the vessel and piping integrity research have demonstrated the high safety margins in scaled vessels and the analytical procedures for calculating vessel behaviour under pressure. Non-destructive examination techniques are being improved. Work is also proceeding to define the material constituents to reduce the susceptibility of irradiation embrittlement and stress corrosion cracking. The thermal-hydraulic tests have covered the various phases of a hypothetical loss of coolant accident (LOCA) and activation of the emergency core cooling system (ECCS). These tests have led to the development of engineering correlations to describe the phenomena to further quantify the safety margins in commercial nuclear power plants. Specifically, this paper presents selected experimental data and analytical predictions from the initial tests in LOFT and SEMISCALE. Comparisons and evaluations are made between the data and analytical predictions. Significant results and conclusions are presented regarding the behaviour of emergency core cooling systems in a LOCA environment: the ability to predict LOCA-type experiments over a scaling range of thirty and the thermal-hydraulic behaviour of components such as pumps in an integral system LOCA environment. The fuel behaviour research has provided valuable information on decay heat, cladding oxidation, fuel rod behaviour and fuel metling. Both the decay heat and the cladding oxidation have been shown to be lower than assumed in the licensing evaluations. The fuel behaviour and thermo-hydraulic research is being integrated into computer codes to be used to provide additional

Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

Energy Technology Data Exchange (ETDEWEB)

Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

2007-06-30

This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition

Twenty-third water reactor safety information meeting: Volume 1, plenary session, high burnup fuel behavior, thermal hydraulic research. Proceedings

Energy Technology Data Exchange (ETDEWEB)

Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

1996-03-01

This three-volume report contains papers presented at the Twenty- Third Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 23-25, 1995. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Italy, Japan, Norway, Russia, Sweden, and Switzerland. This document, Volume 1, present topics on High Burnup Fuel Behavior, Thermal Hydraulic Research, and Plenary Session topics. Individual papers have been cataloged separately.

Twenty-third water reactor safety information meeting: Volume 1, plenary session, high burnup fuel behavior, thermal hydraulic research. Proceedings

International Nuclear Information System (INIS)

Monteleone, S.

1996-03-01

This three-volume report contains papers presented at the Twenty- Third Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 23-25, 1995. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Italy, Japan, Norway, Russia, Sweden, and Switzerland. This document, Volume 1, present topics on High Burnup Fuel Behavior, Thermal Hydraulic Research, and Plenary Session topics. Individual papers have been cataloged separately

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.

RDS - A systematic approach towards system thermal hydraulics input code development for a comprehensive deterministic safety analysis

International Nuclear Information System (INIS)

Salim, Mohd Faiz; Roslan, Ridha; Ibrahim, Mohd Rizal Mamat

2014-01-01

Deterministic Safety Analysis (DSA) is one of the mandatory requirements conducted for Nuclear Power Plant licensing process, with the aim of ensuring safety compliance with relevant regulatory acceptance criteria. DSA is a technique whereby a set of conservative deterministic rules and requirements are applied for the design and operation of facilities or activities. Computer codes are normally used to assist in performing all required analysis under DSA. To ensure a comprehensive analysis, the conduct of DSA should follow a systematic approach. One of the methodologies proposed is the Standardized and Consolidated Reference Experimental (and Calculated) Database (SCRED) developed by University of Pisa. Based on this methodology, the use of Reference Data Set (RDS) as a pre-requisite reference document for developing input nodalization was proposed. This paper shall describe the application of RDS with the purpose of assessing its effectiveness. Two RDS documents were developed for an Integral Test Facility of LOBI-MOD2 and associated Test A1-83. Data and information from various reports and drawings were referred in preparing the RDS. The results showed that by developing RDS, it has made possible to consolidate all relevant information in one single document. This is beneficial as it enables preservation of information, promotes quality assurance, allows traceability, facilitates continuous improvement, promotes solving of contradictions and finally assisting in developing thermal hydraulic input regardless of whichever code selected. However, some disadvantages were also recognized such as the need for experience in making engineering judgments, language barrier in accessing foreign information and limitation of resources. Some possible improvements are suggested to overcome these challenges

RDS - A systematic approach towards system thermal hydraulics input code development for a comprehensive deterministic safety analysis

Energy Technology Data Exchange (ETDEWEB)

Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my [Nuclear Energy Department, Tenaga Nasional Berhad, Level 32, Dua Sentral, 50470 Kuala Lumpur (Malaysia); Roslan, Ridha [Nuclear Installation Division, Atomic Energy Licensing Board, Batu 24, Jalan Dengkil, 43800 Dengkil, Selangor (Malaysia); Ibrahim, Mohd Rizal Mamat [Technical Support Division, Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

2014-02-12

Deterministic Safety Analysis (DSA) is one of the mandatory requirements conducted for Nuclear Power Plant licensing process, with the aim of ensuring safety compliance with relevant regulatory acceptance criteria. DSA is a technique whereby a set of conservative deterministic rules and requirements are applied for the design and operation of facilities or activities. Computer codes are normally used to assist in performing all required analysis under DSA. To ensure a comprehensive analysis, the conduct of DSA should follow a systematic approach. One of the methodologies proposed is the Standardized and Consolidated Reference Experimental (and Calculated) Database (SCRED) developed by University of Pisa. Based on this methodology, the use of Reference Data Set (RDS) as a pre-requisite reference document for developing input nodalization was proposed. This paper shall describe the application of RDS with the purpose of assessing its effectiveness. Two RDS documents were developed for an Integral Test Facility of LOBI-MOD2 and associated Test A1-83. Data and information from various reports and drawings were referred in preparing the RDS. The results showed that by developing RDS, it has made possible to consolidate all relevant information in one single document. This is beneficial as it enables preservation of information, promotes quality assurance, allows traceability, facilitates continuous improvement, promotes solving of contradictions and finally assisting in developing thermal hydraulic input regardless of whichever code selected. However, some disadvantages were also recognized such as the need for experience in making engineering judgments, language barrier in accessing foreign information and limitation of resources. Some possible improvements are suggested to overcome these challenges.

Development of an Owner Engineer's independent capability in NPP safety and licensing

International Nuclear Information System (INIS)

Auglaire, M.; Bayart, D.; D'Eer, A.; Polet, F.; Vanhoenacker, L.; Zhang, J.

2002-01-01

As Owner's Engineer to Electrabel, the Belgian utility which owns and operates the 7 NPPs in Belgium, Tractebel Energy Engineering has gained considerable experience in the field of ten-yearly safety overhauls of NPPs since 1983. It has developed a methodology leading to proposing corrective actions by means of a global and integrated approach in which safety improvement costs are optimized. Safety issues addressed during those projects encompass the writing of Probabilistic Safety Assessment studies, post-TMI recommendations implementation, the installation of autocatalytic recombiners, accident studies, protection against pressurized thermal shock, impact of flooding of internal or external origin, implementation of severe accident management guidelines, re-evaluation of the environment, verification of extreme climate conditions, updating of the Safety Analysis Reports, operation review. (author)

Access Safety Systems – New Concepts from the LHC Experience

CERN Document Server

Ladzinski, T; di Luca, S; Hakulinen, T; Hammouti, L; Riesco, T; Nunes, R; Ninin, P; Juget, J-F; Havart, F; Valentini, F; Sanchez-Corral Mena, E

2011-01-01

The LHC Access Safety System has introduced a number of new concepts into the domain of personnel protection at CERN. These can be grouped into several categories: organisational, architectural and concerning the end-user experience. By anchoring the project on the solid foundations of the IEC 61508/61511 methodology, the CERN team and its contractors managed to design, develop, test and commission on time a SIL3 safety system. The system uses a successful combination of the latest Siemens redundant safety programmable logic controllers with a traditional relay logic hardwired loop. The external envelope barriers used in the LHC include personnel and material access devices, which are interlocked door-booths introducing increased automation of individual access control, thus removing the strain from the operators. These devices ensure the inviolability of the controlled zones by users not holding the required credentials. To this end they are equipped with personnel presence detectors and th...

Adsorption thermal energy storage for cogeneration in industrial batch processes: Experiment, dynamic modeling and system analysis

International Nuclear Information System (INIS)

Schreiber, Heike; Graf, Stefan; Lanzerath, Franz; Bardow, André

2015-01-01

Adsorption thermal energy storage is investigated for heat supply with cogeneration in industrial batch processes. The feasibility of adsorption thermal energy storage is demonstrated with a lab-scale prototype. Based on these experiments, a dynamic model is developed and successfully calibrated to measurement data. Thereby, a reliable description of the dynamic behavior of the adsorption thermal energy storage unit is achieved. The model is used to study and benchmark the performance of adsorption thermal energy storage combined with cogeneration for batch process energy supply. As benchmark, we consider both a peak boiler and latent thermal energy storage based on a phase change material. Beer brewing is considered as an example of an industrial batch process. The study shows that adsorption thermal energy storage has the potential to increase energy efficiency significantly; primary energy consumption can be reduced by up to 25%. However, successful integration of adsorption thermal storage requires appropriate integration of low grade heat: Preferentially, low grade heat is available at times of discharging and in demand when charging the storage unit. Thus, adsorption thermal energy storage is most beneficial if applied to a batch process with heat demands on several temperature levels. - Highlights: • A highly efficient energy supply for industrial batch processes is presented. • Adsorption thermal energy storage (TES) is analyzed in experiment and simulation. • Adsorption TES can outperform both peak boilers and latent TES. • Performance of adsorption TES strongly depends on low grade heat temperature.

Safety; Avertissement

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-07-01

This annual report of the Senior Inspector for the Nuclear Safety, analyses the nuclear safety at EDF for the year 1999 and proposes twelve subjects of consideration to progress. Five technical documents are also provided and discussed concerning the nuclear power plants maintenance and safety (thermal fatigue, vibration fatigue, assisted control and instrumentation of the N4 bearing, 1300 MW reactors containment and time of life of power plants). (A.L.B.)

Fusion safety codes International modeling with MELCOR and ATHENA- INTRA

CERN Document Server

Marshall, T; Topilski, L; Merrill, B

2002-01-01

For a number of years, the world fusion safety community has been involved in benchmarking their safety analyses codes against experiment data to support regulatory approval of a next step fusion device. This paper discusses the benchmarking of two prominent fusion safety thermal-hydraulic computer codes. The MELCOR code was developed in the US for fission severe accident safety analyses and has been modified for fusion safety analyses. The ATHENA code is a multifluid version of the US-developed RELAP5 code that is also widely used for fusion safety analyses. The ENEA Fusion Division uses ATHENA in conjunction with the INTRA code for its safety analyses. The INTRA code was developed in Germany and predicts containment building pressures, temperatures and fluid flow. ENEA employs the French-developed ISAS system to couple ATHENA and INTRA. This paper provides a brief introduction of the MELCOR and ATHENA-INTRA codes and presents their modeling results for the following breaches of a water cooling line into the...

Experience with nuclear safety standards development in non-governmental international organizations

International Nuclear Information System (INIS)

Becker, K.

1985-01-01

Besides the IAEA as a 'governmental' organization dealing with basic safety recommendations addressed primarily to the national regulatory bodies in developing countries, two closely related non-governmental international standards organizations have gained extensive experience in the field of nuclear standardization. Over more than 25 years since their formation, both (a) the International Organization for Standardization's (ISO) Technical Committee 85 'Nuclear Energy', in particular in its Sub-Committee 3 'Reactor Technology and Safety' and (b) the International Electrotechnical Commission's (IEC) Technical Committee 45 'Nuclear Instrumentation' have published numerous standards. A brief review is given of these, draft standards, and other documents planned to become international standards. Many of them deal with rather specialized topics typical for 'industrial' standards such as standardized procedures, instruments, methods, materials, test methods, terminology, and signs and symbols, but others are directly related to more basic safety issues. In some areas such as quality assurance, seismic aspects of siting and terminology, there has been in the past occasional overlap in the activities of the NUSS programme, IEC and ISO. Letters of Understanding have since 1981 contributed to clarifying the borderlines and to avoiding redundant efforts. Also, some experiences and problems are described arising, for example, from the harmonization of different national safety philosophies and traditions into universally accepted international standards, and the transfer of international standards into national standards systems. Finally, based on a recent comprehensive compilation of some 3300 nuclear standards and standards projects, an attempt is made to present a cost/benefit analysis and an outlook on future developments. (author)

Impact of LMFBR operating experience on PFBR design

International Nuclear Information System (INIS)

Bhoje, S.B.; Chetal, S.C.; Chellapandi, P.; Govindarajan, S.; Lee, S.M.; Kameswara Rao, A.S.L.; Prabhakar, R.; Raghupathy, S.; Sodhi, B.S.; Sundaramoorthy, T.R.; Vaidyanathan, G.

2000-01-01

PFBR is a 500 MWe, sodium cooled, pool type, fast breeder reactor currently under detailed design. It is essential to reduce the capital cost of PFBR in order to make it competitive with thermal reactors. Operating experience of LMFBRs provides a vital input towards simplification of the design, improving its reliability, enhancing safety and achieving overall cost reduction. This paper includes a summary of LMFBR operating experience and details the design features of PFBR as influenced by operating experience of LMFBRs. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-09-01

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

Study on Tensile Fatigue Behavior of Thermal Butt Fusion in Safety Class III High-Density Polyethylene Buried Piping in Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Kim, Jong Sung; Lee, Young Ju [Sunchon National University, Suncheon (Korea, Republic of); Oh, Young Jin [KEPCO E and C, Yongin (Korea, Republic of)

2015-01-15

High-density polyethylene (HDPE) piping, which has recently been applied to safety class III piping in nuclear power plants, can be butt-joined through the thermal fusion process, which heats two fused surfaces and then subject to axial pressure. The thermal fusion process generates bead shapes on the butt fusion. The stress concentrations caused by the bead shapes may reduce the fatigue lifetime. Thus, investigating the effect of the thermal butt fusion beads on fatigue behavior is necessary. This study examined the fatigue behavior of thermal butt fusion via a tensile fatigue test under stress-controlled conditions using finite element elastic stress analysis. Based on the results, the presence of thermal butt fusion beads was confirmed to reduce the fatigue lifetime in the low-cycle fatigue region while having a negligible effect in the medium- and high-cycle fatigue regions.

Study on Tensile Fatigue Behavior of Thermal Butt Fusion in Safety Class III High-Density Polyethylene Buried Piping in Nuclear Power Plants

International Nuclear Information System (INIS)

Kim, Jong Sung; Lee, Young Ju; Oh, Young Jin

2015-01-01

High-density polyethylene (HDPE) piping, which has recently been applied to safety class III piping in nuclear power plants, can be butt-joined through the thermal fusion process, which heats two fused surfaces and then subject to axial pressure. The thermal fusion process generates bead shapes on the butt fusion. The stress concentrations caused by the bead shapes may reduce the fatigue lifetime. Thus, investigating the effect of the thermal butt fusion beads on fatigue behavior is necessary. This study examined the fatigue behavior of thermal butt fusion via a tensile fatigue test under stress-controlled conditions using finite element elastic stress analysis. Based on the results, the presence of thermal butt fusion beads was confirmed to reduce the fatigue lifetime in the low-cycle fatigue region while having a negligible effect in the medium- and high-cycle fatigue regions

Pressurized-thermal-shock experiments: PTSE-1 results and PTSE-2 plans

International Nuclear Information System (INIS)

Bryan, R.H.; Nanstad, R.K.; Wanner, R.; Merkle, J.G.; Robinson, G.C.; Whitman, G.D.

1985-01-01

The first pressurized-thermal-shock experiment (PTSE-1) was performed with a vessel with a 1-m-long flaw in a plug of specially tempered steel having the composition of SA-508 forging steel. The second experiment (PTSE-2) will have a similar arrangement, but the material in which the flaw will be implanted is being prepared to have low tearing resistance. Special tempering of a 2 1/4 Cr - 1 Mo steel plate has been shown to induce a low Charpy impact energy in the upper-shelf temperature range. The purpose of PTSE-2 is to investigate the fracture behavior of low-upper-shelf material in a vessel under the combined loading of concurrent pressure and thermal shock. The primary objective of the experimental plan is to induce a rapidly propagating cleavage fracture under conditions that are likely to induce a ductile tearing instability at the time of arrest of the cleavage fracture. The secondary objective of the test is to extend the range of the investigation of warm prestressing. 11 figs

INMACS: Operating experience of a mature, computer-assisted control system for nuclear material inventory and criticality safety

International Nuclear Information System (INIS)

Ross, A.M.

1983-01-01

This paper describes the operating experience of INMACS, the Integrated Nuclear Material Accounting and Control System used in the Recycle Fuel Fabrication Laboratories at Chalk River. Since commissioning was completed in 1977, INMACS has checked and recorded approximately 3000 inventory-related transactions involved in fabricating thermal-recycle fuels of (U,Pu)0 2 and (Th,Pu)0 2 . No changes have been necessary to INMACS programs that are used by laboratory staff when moving or processing nuclear material. The various utility programs have allowed efficient management and surveillance of the INMACS data base. Hardware failures and the nuisance of system unavailability at the laboratory terminals have been minimized by regular preventative maintenance. The original efforts in the design and rigorous testing of programs have helped INMACS to be accepted enthusiastically by old and new staff of the laboratories. The work required for nuclear material inventory control is done efficiently and in an atmosphere of safety

Thermal hydraulic analysis of aggressive secondary cooldown in a small break loss of coolant accident with a total loss of high pressure safety injection

International Nuclear Information System (INIS)

Han, Seok Jung; Lim, Ho Gon; Yang, Joon Eon

2003-01-01

To support the development of a Probabilistic Safety Assessment (PSA) model usable in Riskinformed Applications (RIA) for Korea Standard Nuclear power Plants (KSNP), we have performed a thermal hydraulic analysis of Aggressive Secondary Cooldown (ASC) in a 2-inch Small Break Loss Of Coolant Accident (SBLOCA) with a total loss of High Pressure Safety Injection (HPSI). The present study focuses on the estimation of the success criteria of ASC, and the enhanced understanding of the detailed thermal hydraulic behavior and phenomena. The results have shown that the Reactor Coolant System (RCS) pressure can be reduced to the Low Pressure Safety Injection (LPSI) operation conditions without core damage. It was also shown that more relaxed success criteria compared to those in the previous PSA models of KSNP could be used in the new PSA model. However, it was found that the results could be affected by various parameters related with ASC operation, i.e., reference temperature for the calculation of the cooldown rate and its control method

Strain measurements during pressurized thermal shock experiment

International Nuclear Information System (INIS)

Tarso Vida Gomes, P. de; Julio Ricardo Barreto Cruz; Tanius Rodrigues Mansur; Denis Henrique Bianchi Scaldaferri; Miguel Mattar Neto

2005-01-01

For the life extension of nuclear power plants, the residual life of most of their components must be evaluated along all their operating time. Concerning the reactor pressure vessel, the pressurized thermal shock (PTS) is a very important event to be considered. For better understanding the effects of this kind of event, tests are made. The approach described here consisted of building a simplified in-scale physical model of the reactor pressure vessel, submitting it to the actual operating temperature and pressure conditions and provoking a thermal shock by means of cold water flow in its external surface. To conduct such test, the Nuclear Technology Development Center (CDTN) has been conducting several studies related to PTS and has also built a laboratory that has made possible the simulation of the PTS loading conditions. Several cracks were produced in the external surface of the reactor pressure vessel model. Strain gages were fixed by means of electrical discharge welding over the cracks regions in both external and internal surfaces. The temperature was monitored in 10 points across the vessel wall. The internal pressure was manually controlled and monitored using a pressure transducer. Two PTS experiments were conducted and this paper presents the strain measurement procedures applied to the reactor pressure vessel model, during the PTS, using strain gages experimental methodology. (authors)

Nuclear reactor thermal hydraulics safety analysis and thoughts on FUKUSHIMA

International Nuclear Information System (INIS)

Ninokata, Hisashi

2012-01-01

The first part of this article is to show my thoughts on the accident at Fukushima Daiichi Nuclear Power Station. It is cited from a summary of my lecture talk in Indonesia, in the beginning of the last December, 2011. This talk was based on my previous lecture and seminar talks including those delivered at MIT, June 16, at the ANS Annual Meeting in Hollywood, Florida, June 28 at NURETH-13 in Toronto, September 27, and others. The content is based on the open and latest information available to date in Japan. It may contain some erroneous or uncertain information. I tried to minimize it to my best capability. Also I tried to eliminate any critical issues or opinions that may jeopardize some people who were involved in. The latter half of this article will be excerpts of my recent R and D activities related to the safety-by-design for sodium cooled fast reactors and light water reactors, thermal hydraulics analysis focusing on the simulation-based technology, in particular subchannel analysis and computational fluid dynamics. (J.P.N.)

Analysis of the impact of correlated benchmark experiments on the validation of codes for criticality safety analysis

International Nuclear Information System (INIS)

Bock, M.; Stuke, M.; Behler, M.

2013-01-01

The validation of a code for criticality safety analysis requires the recalculation of benchmark experiments. The selected benchmark experiments are chosen such that they have properties similar to the application case that has to be assessed. A common source of benchmark experiments is the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments' (ICSBEP Handbook) compiled by the 'International Criticality Safety Benchmark Evaluation Project' (ICSBEP). In order to take full advantage of the information provided by the individual benchmark descriptions for the application case, the recommended procedure is to perform an uncertainty analysis. The latter is based on the uncertainties of experimental results included in most of the benchmark descriptions. They can be performed by means of the Monte Carlo sampling technique. The consideration of uncertainties is also being introduced in the supplementary sheet of DIN 25478 'Application of computer codes in the assessment of criticality safety'. However, for a correct treatment of uncertainties taking into account the individual uncertainties of the benchmark experiments is insufficient. In addition, correlations between benchmark experiments have to be handled correctly. For example, these correlations can arise due to different cases of a benchmark experiment sharing the same components like fuel pins or fissile solutions. Thus, manufacturing tolerances of these components (e.g. diameter of the fuel pellets) have to be considered in a consistent manner in all cases of the benchmark experiment. At the 2012 meeting of the Expert Group on 'Uncertainty Analysis for Criticality Safety Assessment' (UACSA) of the OECD/NEA a benchmark proposal was outlined that aimed for the determination of the impact on benchmark correlations on the estimation of the computational bias of the neutron multiplication factor (k eff ). The analysis presented here is based on this proposal. (orig.)

Mineralogical control on thermal damage and the presence of a thermal Kaiser effect during temperature-cycling experiments

Science.gov (United States)

Browning, J.; Daoud, A.; Meredith, P. G.; Mitchell, T. M.

2017-12-01

Volcanic and geothermal systems are in part controlled by the mechanical and thermal stresses acting on them and so it is important to understand the response of volcanic rocks to thermo-mechanical loading. One such response is the well-known `Kaiser stress-memory' effect observed under cyclic mechanical loading. By contrast, the presence of an analogous `Kaiser temperature-memory effect' during cyclic thermal loading has received little attention. We have therefore explored the possibility of a Kaiser temperature-memory effect using three igneous rocks of different composition, grain size and origin; Slaufrudalur Granophyre (SGP), Nea Kameni Andesite (NKA) and Seljadalur Basalt (SB). We present results from a series of thermal stressing experiments in which acoustic emissions (AE) were recorded contemporaneously with changing temperature. Samples of each rock were subjected to both a single heating and cooling cycle to a maximum temperature of 900 °C and multiple heating/cooling cycles to peak temperatures of 350°C, 500°C, 700°C and 900 °C (all at a constant rate of 1°C/min on heating and a natural cooling rate of memory effect in SGP, but not in either NKA and SB. We further find that the vast majority of thermal crack damage is generated upon cooling in the finer grained materials (NKA and SB), but that substantial thermal crack damage is generated during heating in the coarser grained SGP. The total amount of crack damage generated due to heating or cooling is dependent on the mineral composition and, most importantly, the grain size and arrangement, as well as the maximum temperature to which the rock is exposed. Knowledge of thermal stress history and the presence of a Kaiser temperature-memory effect is potentially important in understanding magma chamber dynamics, where the cyclic nature of mechanical and thermal inflation and deflation can lead to sequential accumulation of damage, potentially leading to critical rupture.

INL Experimental Program Roadmap for Thermal Hydraulic Code Validation

Energy Technology Data Exchange (ETDEWEB)

Glenn McCreery; Hugh McIlroy

2007-09-01

Advanced computer modeling and simulation tools and protocols will be heavily relied on for a wide variety of system studies, engineering design activities, and other aspects of the Next Generation Nuclear Power (NGNP) Very High Temperature Reactor (VHTR), the DOE Global Nuclear Energy Partnership (GNEP), and light-water reactors. The goal is for all modeling and simulation tools to be demonstrated accurate and reliable through a formal Verification and Validation (V&V) process, especially where such tools are to be used to establish safety margins and support regulatory compliance, or to design a system in a manner that reduces the role of expensive mockups and prototypes. Recent literature identifies specific experimental principles that must be followed in order to insure that experimental data meet the standards required for a “benchmark” database. Even for well conducted experiments, missing experimental details, such as geometrical definition, data reduction procedures, and manufacturing tolerances have led to poor Benchmark calculations. The INL has a long and deep history of research in thermal hydraulics, especially in the 1960s through 1980s when many programs such as LOFT and Semiscle were devoted to light-water reactor safety research, the EBRII fast reactor was in operation, and a strong geothermal energy program was established. The past can serve as a partial guide for reinvigorating thermal hydraulic research at the laboratory. However, new research programs need to fully incorporate modern experimental methods such as measurement techniques using the latest instrumentation, computerized data reduction, and scaling methodology. The path forward for establishing experimental research for code model validation will require benchmark experiments conducted in suitable facilities located at the INL. This document describes thermal hydraulic facility requirements and candidate buildings and presents examples of suitable validation experiments related

YALINA-Thermal Facility Experiments

International Nuclear Information System (INIS)

Kiyavitskaya, H.; Bournos, V.; Fokov, Y.; Routkovskaya, C.; Sadovich, S.; Cintas, A.; Márquez Damián, J.I.; Lopasso, E.M.; Maiorino, J.R.; Carluccio, T.; Rossi, P.C.R.; Antunes, A.; Oliveira, F.L. de; Lee, S.M.; Xia, P.; Shi, Y.; Xia, H.; Zhu, Q.; Yu, T.; Wu, X.; Zhang, W.; Cao, J.; Luo, H.; Quan, Y.; Kulkarni, K.; Yadav, R.D.S.; Bajpai, A.; Degweker, S.B.; Modak, R.S.; Park, H.J.; Shim, H.J.; Kim, C.H.; Wojciechowski, A.; Zuta, M.; Pešić, M.; Avramović, I.; Beličev, P.; Gohar, Y.; Talamo, A.; Aliberti, G.

2017-01-01

This Section discussed the results obtained by the Member States participating in the IAEA coordinated research project on Analytical and Experimental Benchmark Analysis on Accelerator Driven Systems, and Low Enriched Uranium Fuel Utilization in Accelerator Driven Subcritical Assembly Systems for the YALINA Thermal facility. Member States used both Monte Carlo and deterministic computational tools to analyse the YALINA Thermal subcritical assembly, including: MCNP5, MCNPX, McCARD, PARTISN, and ERANOS computer programs. All calculations have been performed using the ENDF/B-VI (different modes) nuclear data libraries with the exception of Republic of Korea which used the ENDF/B-VII.0 nuclear data library. Generally, there is a good agreement between the results obtained by all the Member States. Deterministic codes perform space, energy, and angle discretization and materials homogenizations, which introduce approximations affecting the obtained results. In subcritical assemblies, the neutron multiplication and the detector counting rate depend strongly on the external neutron source. Cf and D-D sources provide similar results since they emit neutrons with similar average energy. D-T neutrons trigger (n,xn) reactions and have a longer mean free path, which increases the neutron leakage if the geometry dimensions of the assembly are small, as in the case of the YALINA-Thermal subcritical assembly. Close to criticality, the effect of the external neutron source diminishes since fission neutrons dominate the neutron population.

Lessons from feedback of safety operating experience for reactor physics

International Nuclear Information System (INIS)

Suchomel, J.; Rapavy, S.

1999-01-01

Analyses of events in WWER operations as a part of safety experience feedback provide a valuable source of lessons for reactor physics. Examples of events from Bohunice operation will be shown such as events with inadequate approach to criticality, positive reactivity insertions, expulsion of a control rod from shut-down reactor, problems with reactor protection system and control rods. (Authors)

Design of 6 Mev linear accelerator based pulsed thermal neutron source: FLUKA simulation and experiment

Energy Technology Data Exchange (ETDEWEB)

Patil, B.J., E-mail: bjp@physics.unipune.ac.in [Department of Physics, University of Pune, Pune 411 007 (India); Chavan, S.T.; Pethe, S.N.; Krishnan, R. [SAMEER, IIT Powai Campus, Mumbai 400 076 (India); Bhoraskar, V.N. [Department of Physics, University of Pune, Pune 411 007 (India); Dhole, S.D., E-mail: sanjay@physics.unipune.ac.in [Department of Physics, University of Pune, Pune 411 007 (India)

2012-01-15

The 6 MeV LINAC based pulsed thermal neutron source has been designed for bulk materials analysis. The design was optimized by varying different parameters of the target and materials for each region using FLUKA code. The optimized design of thermal neutron source gives flux of 3 Multiplication-Sign 10{sup 6}ncm{sup -2}s{sup -1} with more than 80% of thermal neutrons and neutron to gamma ratio was 1 Multiplication-Sign 10{sup 4}ncm{sup -2}mR{sup -1}. The results of prototype experiment and simulation are found to be in good agreement with each other. - Highlights: Black-Right-Pointing-Pointer The optimized 6 eV linear accelerator based thermal neutron source using FLUKA simulation. Black-Right-Pointing-Pointer Beryllium as a photonuclear target and reflector, polyethylene as a filter and shield, graphite as a moderator. Black-Right-Pointing-Pointer Optimized pulsed thermal neutron source gives neutron flux of 3 Multiplication-Sign 10{sup 6} n cm{sup -2} s{sup -1}. Black-Right-Pointing-Pointer Results of the prototype experiment were compared with simulations and are found to be in good agreement. Black-Right-Pointing-Pointer This source can effectively be used for the study of bulk material analysis and activation products.

11. international topical meeting on nuclear reactor thermal-hydraulics (NURETH-11)

International Nuclear Information System (INIS)

Lemonnier, H.

2005-01-01

; aerosol transport, deposition and re-entrainment; steam generators thermal-hydraulics; system codes development and assessment; uncertainties analysis; diffuse interface methods and interface tracking methods; C - severe accidents and fires: molten core natural convection and physico-chemical phenomena, modeling and experiments; fuel coolant interaction, modeling and experiments; debris bed cooling; combustion and fires, modeling and experiments; molten corium concrete interaction; D - advanced code developments: fast transient modelling and experiments; multidimensional single-phase or two-phase flow and heat transfer modeling; neutronics and thermal-hydraulics coupling; fluid and structures mechanical interactions; coupled thermal-hydraulics of fluids and structures; thermal-hydraulic dependent corrosion and ablation; E - operation and safety of existing reactors: instabilities and nonlinear dynamics; NPP transients and accidents analysis; RBMK and VVER safety analysis, including the OECD benchmark; F - experimental thermal-hydraulics: boiling heat transfer; CHF and post-CHF heat transfer; condensation heat transfer; integral testing; vibrations, wear and thermal fatigue phenomena; fuel design and performance; G - advanced reactors thermal-hydraulics (gen IV, INPRO, fusion, hydrogen production): accelerator driven reactors; advanced pressurized water reactors thermal-hydraulics; gas cooled fast reactors; gas cooled high temperature reactors; lead and lead-bismuth cooled reactors; future and existing sodium cooled reactors; molten salt reactors; H - waste management thermal-hydraulics: thermal-hydraulics problems related to waste processing and storage; I - thermal-hydraulics of non electricity generating nuclear equipment: sono-fusion (cavitation induced bubble fusion; hydrogen producing nuclear reactors

BOW SHOCK FRAGMENTATION DRIVEN BY A THERMAL INSTABILITY IN LABORATORY ASTROPHYSICS EXPERIMENTS

Energy Technology Data Exchange (ETDEWEB)

Suzuki-Vidal, F.; Lebedev, S. V.; Pickworth, L. A.; Swadling, G. F.; Skidmore, J.; Hall, G. N.; Bennett, M.; Bland, S. N.; Burdiak, G.; De Grouchy, P.; Music, J.; Suttle, L. [Blackett Laboratory, Imperial College London, Prince Consort Road, London SW7 2BW (United Kingdom); Ciardi, A. [Sorbonne Universités, UPMC Univ. Paris 6, UMR 8112, LERMA, F-75005, Paris (France); Rodriguez, R.; Gil, J. M.; Espinosa, G. [Departamento de Fisica de la Universidad de Las Palmas de Gran Canaria, E-35017 Las Palmas de Gran Canaria (Spain); Hartigan, P. [Department of Physics and Astronomy, Rice University, 6100 S. Main, Houston, TX 77521-1892 (United States); Hansen, E.; Frank, A., E-mail: f.suzuki@imperial.ac.uk [Department of Physics and Astronomy, University of Rochester, Rochester, NY 14627 (United States)

2015-12-20

The role of radiative cooling during the evolution of a bow shock was studied in laboratory-astrophysics experiments that are scalable to bow shocks present in jets from young stellar objects. The laboratory bow shock is formed during the collision of two counterstreaming, supersonic plasma jets produced by an opposing pair of radial foil Z-pinches driven by the current pulse from the MAGPIE pulsed-power generator. The jets have different flow velocities in the laboratory frame, and the experiments are driven over many times the characteristic cooling timescale. The initially smooth bow shock rapidly develops small-scale nonuniformities over temporal and spatial scales that are consistent with a thermal instability triggered by strong radiative cooling in the shock. The growth of these perturbations eventually results in a global fragmentation of the bow shock front. The formation of a thermal instability is supported by analysis of the plasma cooling function calculated for the experimental conditions with the radiative packages ABAKO/RAPCAL.

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

In-flight thermal experiments for LISA Pathfinder: Simulating temperature noise at the Inertial Sensors

International Nuclear Information System (INIS)

Armano, M; Audley, H; Born, M; Danzmann, K; Diepholz, I; Auger, G; Binetruy, P; Baird, J; Bortoluzzi, D; Brandt, N; Fitzsimons, E; Bursi, A; Caleno, M; Cavalleri, A; Cesarini, A; Dolesi, R; Ferroni, V; Cruise, M; Dunbar, N; Ferraioli, L

2015-01-01

Thermal Diagnostics experiments to be carried out on board LISA Pathfinder (LPF) will yield a detailed characterisation of how temperature fluctuations affect the LTP (LISA Technology Package) instrument performance, a crucial information for future space based gravitational wave detectors as the proposed eLISA. Amongst them, the study of temperature gradient fluctuations around the test masses of the Inertial Sensors will provide as well information regarding the contribution of the Brownian noise, which is expected to limit the LTP sensitivity at frequencies close to 1 mHz during some LTP experiments. In this paper we report on how these kind of Thermal Diagnostics experiments were simulated in the last LPF Simulation Campaign (November, 2013) involving all the LPF Data Analysis team and using an end-to-end simulator of the whole spacecraft. Such simulation campaign was conducted under the framework of the preparation for LPF operations. (paper)

Drug research methodology. Volume 1, The alcohol-highway safety experience and its applicability to other drugs

Science.gov (United States)

1980-03-01

This report presents the findings of a workshop concerning the alcohol and highway safety experience, which includes research efforts to define the drinking-driving problem and societal responses to reduce the increased highway safety risk attributab...

MHTGR: New production reactor summary of experience base

International Nuclear Information System (INIS)

1988-03-01

Worldwide interest in the Modular High-Temperature Gas-Cooled Reactor (MHTGR) stems from the capability of the system to retain the advanced fuel and thermal performance while providing unparalleled levels of safety. The small power level of the MHTGR and its passive systems give it a margin of safety not attained by other concepts being developed for power generation. This report covers the experience base for the key nuclear system, components, and processes related to the MHTGR-NPR. 9 refs., 39 figs., 9 tabs

Computational Fluid Dynamics (CFD) in Nuclear Reactor Safety (NRS) - Proceedings of the workshop on Experiments and CFD Code Application to Nuclear Reactor Safety (XCFD4NRS)

International Nuclear Information System (INIS)

2008-01-01

Computational Fluid Dynamics (CFD) is to an increasing extent being adopted in nuclear reactor safety analyses as a tool that enables specific safety relevant phenomena occurring in the reactor coolant system to be better described. The Committee on the Safety of Nuclear Installations (CSNI), which is responsible for the activities of the OECD Nuclear Energy Agency that support advancing the technical base of the safety of nuclear installations, has in recent years conducted an important activity in the CFD area. This activity has been carried out within the scope of the CSNI working group on the analysis and management of accidents (GAMA), and has mainly focused on the formulation of user guidelines and on the assessment and verification of CFD codes. It is in this GAMA framework that a first workshop CFD4NRS was organized and held in Garching, Germany in 2006. Following the CFD4NRS workshop, this XCFD4NRS Workshop was intended to extend the forum created for numerical analysts and experimentalists to exchange information in the field of Nuclear Reactor Safety (NRS) related activities relevant to Computational Fluid Dynamics (CFD) validation, but this time with more emphasis placed on new experimental techniques and two-phase CFD applications. The purpose of the workshop was to provide a forum for numerical analysts and experimentalists to exchange information in the field of NRS-related activities relevant to CFD validation, with the objective of providing input to GAMA CFD experts to create a practical, state-of-the-art, web-based assessment matrix on the use of CFD for NRS applications. The scope of XCFD4NRS includes: - Single-phase and two-phase CFD simulations with an emphasis on validation in areas such as: boiling flows, free-surface flows, direct contact condensation and turbulent mixing. These applications should relate to NRS-relevant issues such as: pressurized thermal shocks, critical heat flux, pool heat exchangers, boron dilution, hydrogen

EXPERIENCE NETWORKING UNIVERSITY OF EDUCATION TRAINING MASTERS SAFETY OF LIFE

Directory of Open Access Journals (Sweden)

Elvira Mikhailovna Rebko

2016-02-01

Full Text Available The article discloses experience networking of universities (Herzen State Pedagogical University and Sakhalin State University in the development and implementation of joint training programs for master’s education in the field of life safety «Social security in the urban environment». The novelty of the work is to create a schematic design of basic educational training program for master’s education in the mode of networking, and to identify effective instructional techniques and conditions of networking.Purpose – present the results of the joint development of a network of the basic educational program (BEP, to identify the stages of networking, to design a generalized scheme of development and implementation of a network of educational training program for master’s education in the field of life safety.Results generalized model of networking partner institutions to develop and implement the basic educational program master.Practical implications: the education process for Master of Education in the field of health and safety in Herzen State Pedagogical University and Sakhalin State University.

Looking back on safety management in construction of advanced thermal nuclear power plant 'Fugen'

International Nuclear Information System (INIS)

Emori, Kengo

1979-01-01

The safety management of the advanced thermal nuclear power plant ''Fugen'' during the period from the preparation of plant construction in October, 1970, to the full power operation in March, 1979, is looked back and explained. Any large human and material accidents did not occur during the long construction time. The total numbers of persons and hours were 1.397 x 10 6 workers and 11.55 x 10 6 hours, respectively. The number of labor accidents was twenty with no dead person, the number of loss days was 645 days, the number of accident rate was 1.73, the intensity rate was 0.06, and the mean rate of labor accidents per year per 1000 workers was 6.4. The radiation exposure dose was 65.27 man-rem for the managed 1804 workers in total, during the testing and operating periods. These data show that the safety management for ''Fugen'' is very excellent, considering the following special features: 1) there were many works which were carried out for the first time, 2) the construction of the plant was conducted by five contractors taking partial charge, there were many kinds of construction works, the construction and testing periods were long, and the workers had to go to the site from Tsuruga city by car and bus. The organization of preventing disasters, the concrete implementation items for safety management, including the planning of activities, various meetings, patrol, education and training, the honoring system, the prevention of traffic accident and so on, and the results of actual safety management are explained with the reflection. (Nakai, Y.)

What is the best clothing to prevent heat and cold stress? Experiences with thermal manikin.

Science.gov (United States)

Magyar, Z; Tamas, R

2013-02-01

The present study summarizes the current knowledge of the heat and cold stress which might significantly affect military activities and might also occur among travellers who are not well adapted to weather variations during their journey. The selection of the best clothing is a very important factor in preserving thermal comfort. Our experiences with thermal manikin are also represented in this paper.

Experimental study on thermal-hydraulic behaviors of a pressure balanced coolant injection system for a passive safety light water reactor JPSR

Energy Technology Data Exchange (ETDEWEB)

Satoh, Takashi; Watanabe, Hironori; Araya, Fumimasa; Nakajima, Katsutoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iwamura, Takamichi; Murao, Yoshio

1998-02-01

A conceptual design study of a passive safety light water reactor JPSR has been performed at Japan Atomic Energy Research Institute JAERI. A pressure balanced coolant injection experiment has been carried out, with an objective to understand thermal-hydraulic characteristics of a passive coolant injection system which has been considered to be adopted to JPSR. This report summarizes experimental results and data recorded in experiment run performed in FY. 1993 and 1994. Preliminary experiments previously performed are also briefly described. As the results of the experiment, it was found that an initiation of coolant injection was delayed with increase in a subcooling in the pressure balance line. By inserting a separation device which divides the inside of core make-up tank (CMT) into several small compartments, a diffusion of a high temperature region formed just under the water surface was restrained and then a steam condensation was suppressed. A time interval from an uncovery of the pressure balance line to the initiation of the coolant injection was not related by a linear function with a discharge flow rate simulating a loss-of-coolant accident (LOCA) condition. The coolant was injected intermittently by actuation of a trial fabricated passive valve actuated by pressure difference for the present experiment. It was also found that the trial passive valve had difficulties in setting an actuation set point and vibrations noises and some fraction of the coolant was remained in CMT without effective use. A modification was proposed for resolving these problems by introducing an anti-closing mechanism. (author)

RDS; A systematic approach towards system thermal hydraulics input code development for a comprehensive deterministic safety analysis

International Nuclear Information System (INIS)

Mohd Faiz Salim; Ridha Roslan; Mohd Rizal Mamat

2013-01-01

Full-text: Deterministic Safety Analysis (DSA) is one of the mandatory requirements conducted for Nuclear Power Plant licensing process, with the aim of ensuring safety compliance with relevant regulatory acceptance criteria. DSA is a technique whereby a set of conservative deterministic rules and requirements are applied for the design and operation of facilities or activities. Computer codes are normally used to assist in performing all required analysis under DSA. To ensure a comprehensive analysis, the conduct of DSA should follow a systematic approach. One of the methodologies proposed is the Standardized and Consolidated Reference Experimental (and Calculated) Database (SCRED) developed by University of Pisa. Based on this methodology, the use of Reference Data Set (RDS) as a pre-requisite reference document for developing input nodalization was proposed. This paper shall describe the application of RDS with the purpose of assessing its effectiveness. Two RDS documents were developed for an Integral Test Facility of LOBIMOD2 and associated Test A1-83. Data and information from various reports and drawings were referred in preparing the RDS. The results showed that by developing RDS, it has made possible to consolidate all relevant information in one single document. This is beneficial as it enables preservation of information, promotes quality assurance, allows traceability, facilitates continuous improvement, promotes solving of contradictions and finally assisting in developing thermal hydraulic input regardless of whichever code selected. However, some disadvantages were also recognized such as the need for experience in making engineering judgments, language barrier in accessing foreign information and limitation of resources. Some possible improvements are suggested to overcome these challenges. (author)

The labor protection and safety measures at the electrolysis department

International Nuclear Information System (INIS)

Galushkin, N.V.

1995-01-01

This chapter of monograph is devoted to labor protection and safety measures at the electrolysis department. Thus, the characteristics of dangerous and harmful production factors as well as the danger of thermal burns and thermal exposure were considered. Safety requirements on labor safety were studied.

Performance improvement of the Annular Core Pulse Reactor for reactor safety experiments

International Nuclear Information System (INIS)

Reuscher, J.A.; Pickard, P.S.

1976-01-01

The Annular Core Pulse Reactor (ACPR) is a TRIGA type reactor which has been in operation at Sandia Laboratories since 1967. The reactor is utilized in a wide variety of experimental programs which include radiation effects, neutron radiography, activation analysis, and fast reactor safety. During the past several years, the ACPR has become an important experimental facility for the United States Fast Reactor Safety Research Program and questions of interest to the safety of the LMFBR are being addressed. In order to enhance the capabilities of the ACPR for reactor safety experiments, a project to improve the performance of the reactor was initiated. It is anticipated that the pulse fluence can be increased by a factor of 2.0 to 2.5 utilizing a two-region core concept with high heat capacity fuel elements around the central irradiation cavity. In addition, the steady-state power of the reactor will be increased by about a factor of two. The new features of the improvements are described

The changing landscape of thermal experience and warmth in older people’s dwellings

International Nuclear Information System (INIS)

Tweed, Christopher; Humes, Nicholas; Zapata-Lancaster, Gabriela

2015-01-01

The UK's carbon dioxide reduction policy initiatives often treat environmental conditions in buildings as averaged values of air temperature that flatten spatial variations. This discounts the influence of varying thermal conditions on how people use buildings and the impact this may have on energy consumption. This paper explores the intersection between older people's thermal experience, spatial and temporal variations in thermal conditions in a dwelling and the influence this has on occupants' use of space. The paper reports on qualitative studies in homes with both conventional and newly installed low carbon heating systems. The results suggest that older people are sensitive to and adept at exploiting variations in the dynamic ‘landscape’ of warmth to achieve desired thermal preferences and that they modify their dwellings to improve the quality of the thermal environment. There is also some evidence of a ‘spatial rebound’ effect after energy upgrades, when occupants inhabit rooms they previously could not afford to heat. The nature of qualitative research precludes robust recommendations for policy. However, one important avenue to explore further appears to be that householders may be more strongly motivated by interventions offering improvements across a range of aspects rather than on energy savings alone. -- Highlights: •Thermal variations across space affect the use of space within a dwelling. •Older people show interest in and understanding of thermal behavior of dwellings. •Older people pursue the thermal conditions they desire. •Older people take actions to modify the quality of the thermal environment in their house. •Changes made to the house may not be solely motivated by thermal comfort concerns

Safety in the Utilization and Modification of Research Reactors. Specific Safety Guide

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-07-15

This Safety Guide is a revision of Safety Series No. 35-G2 on safety in the utilization and modification of research reactors. It provides recommendations on meeting the requirements for the categorization, safety assessment and approval of research reactor experiments and modification projects. Specific safety considerations in different phases of utilization and modification projects are covered, including the pre-implementation, implementation and post-implementation phases. Guidance is also provided on the operational safety of experiments, including in the handling, dismantling, post-irradiation examination and disposal of experimental devices. Examples of the application of the safety categorization process for experiments and modification projects and of the content of the safety analysis report for an experiment are also provided. Contents: 1. Introduction; 2. Management system for the utilization and modification of a research reactor; 3. Categorization, safety assessment and approval of an experiment or modification; 4. Safety considerations for the design of an experiment or modification; 5. Pre-implementation phase of a modification or utilization project; 6. Implementation phase of a modification or utilization project; 7. Post-implementation phase of a utilization or modification project; 8. Operational safety of experiments at a research reactor; 9. Safety considerations in the handling, dismantling, post-irradiation examination and disposal of experimental devices; 10. Safety aspects of out-of-reactor-core installations; Annex I: Example of a checklist for the categorization of an experiment or modification at a research reactor; Annex II: Example of the content of the safety analysis report for an experiment at a research reactor; Annex III: Examples of reasons for a modification at a research reactor.

Safety in the Utilization and Modification of Research Reactors. Specific Safety Guide

International Nuclear Information System (INIS)

2012-01-01

This Safety Guide is a revision of Safety Series No. 35-G2 on safety in the utilization and modification of research reactors. It provides recommendations on meeting the requirements for the categorization, safety assessment and approval of research reactor experiments and modification projects. Specific safety considerations in different phases of utilization and modification projects are covered, including the pre-implementation, implementation and post-implementation phases. Guidance is also provided on the operational safety of experiments, including in the handling, dismantling, post-irradiation examination and disposal of experimental devices. Examples of the application of the safety categorization process for experiments and modification projects and of the content of the safety analysis report for an experiment are also provided. Contents: 1. Introduction; 2. Management system for the utilization and modification of a research reactor; 3. Categorization, safety assessment and approval of an experiment or modification; 4. Safety considerations for the design of an experiment or modification; 5. Pre-implementation phase of a modification or utilization project; 6. Implementation phase of a modification or utilization project; 7. Post-implementation phase of a utilization or modification project; 8. Operational safety of experiments at a research reactor; 9. Safety considerations in the handling, dismantling, post-irradiation examination and disposal of experimental devices; 10. Safety aspects of out-of-reactor-core installations; Annex I: Example of a checklist for the categorization of an experiment or modification at a research reactor; Annex II: Example of the content of the safety analysis report for an experiment at a research reactor; Annex III: Examples of reasons for a modification at a research reactor.

[Experience feedback committee: a method for patient safety improvement].

Science.gov (United States)

François, P; Sellier, E; Imburchia, F; Mallaret, M-R

2013-04-01

An experience feedback committee (CREX, Comité de Retour d'EXpérience) is a method which contributes to the management of safety of care in a medical unit. Originally used for security systems of civil aviation, the method has been adapted to health care facilities and successfully implemented in radiotherapy units and in other specialties. We performed a brief review of the literature for studies reporting data on CREX established in hospitals. The review was performed using the main bibliographic databases and Google search results. The CREX is designed to analyse incidents reported by professionals. The method includes monthly meetings of a multi-professional committee that reviews the reported incidents, chooses a priority incident and designates a "pilot" responsible for investigating the incident. The investigation of the incident involves a systemic analysis method and a written synthesis presented at the next meeting of the committee. The committee agrees on actions for improvement that are suggested by the analysis and follows their implementation. Systems for the management of health care, including reporting systems, are organized into three levels: the medical unit, the hospital and the country as a triple loop learning process. The CREX is located in the base level, short loop of risk management and allows direct involvement of care professionals in patient safety. Safety of care has become a priority of health systems. In this context, the CREX can be a useful vehicle for the implementation of a safety culture in medical units. Copyright © 2013 Elsevier Masson SAS. All rights reserved.

Phenomenological Studies on Sodium for CSP Applications: A Safety Review

Energy Technology Data Exchange (ETDEWEB)

Armijo, Kenneth Miguel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Concentrating Solar Technologies Dept.; Andraka, Charles E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Concentrating Solar Technologies Dept.

2016-09-01

Sodium as a heat transfer fluid (HTF) can achieve temperatures above 700°C to improve power cycle performance for reducing large infrastructure costs of high-temperature systems. Current concentrating solar power (CSP) sensible HTF’s (e.g. air, salts) have poor thermal conductivity, and thus low heat transfer capabilities, requiring a large receiver. The high thermal conductivity of sodium has demonstrated high heat transfer rates on dish and towers systems, which allow a reduction in receiver area by a factor of two to four, reducing re-radiation and convection losses and cost by a similar factor. Sodium produces saturated vapor at pressures suitable for transport starting at 600°C and reaches one atmosphere at 870°C, providing a wide range of suitable latent operating conditions that match proposed high temperature, isothermal input power cycles. This advantage could increase the receiver and system efficiency while lowering the cost of CSP tower systems. Although there are a number of desirable thermal performance advantages associated with sodium, its propensity to rapidly oxidize presents safety challenges. This investigation presents a literature review that captures historical operations/handling lessons for advanced sodium systems, and the current state-of-knowledge related to sodium combustion behavior. Technical and operational solutions addressing sodium safety and applications in CSP will be discussed, including unique safety hazards and advantages using latent sodium. Operation and maintenance experience from the nuclear industry with sensible and latent systems will also be discussed in the context of safety challenges and risk mitigation solutions.

Low gravity fluid-thermal experiments

International Nuclear Information System (INIS)

Krotiuk, W.J.; Cuta, J.M.

1987-06-01

Pacific Northwest Laboratory (PNL) is the lead laboratory for the thermal-hydraulic research in the US Department of Energy Multimegawatt Space Nuclear Power Program. PNL must provide the tools necessary to analyze proposed space reactor concepts, which include single- and two-phase alkali metal and gas-cooled designs. PNL has divided its activities for this task into three basic areas: computer code development, thermal-hydraulic modeling, and experimentation. The subject of this paper is the low-gravity experimental program currently underway at PNL in support of the MMW Program

Thermal neutron pulsed parameters in non-hydrogenous systems. Experiment for lead grains

International Nuclear Information System (INIS)

Drozdowicz, K.; Gabanska, B.; Kosik, M.; Krynicka, E.; Woznicka, U.; Zaleski, T.

1997-01-01

In Czubek's method of measurement of the thermal neutron macroscopic absorption cross section a two-region geometry is applied where the investigated sample is surrounded by an external moderator. Both regions in the measurements made up till now were hydrogenous, which means the same type of the thermal neutron transport properties. In the paper a theoretical consideration to use non-hydrogenous materials as the samples is presented. Pulsed neutron measurements have been performed on homogeneous material in a geometry of the classic experiment with the variable geometric buckling. Two decay constants have been measured for different cylindrical samples of small lead grains (a lead shot). (author)

Gas-cooled reactor thermal-hydraulics using CAST3M and CRONOS2 codes

International Nuclear Information System (INIS)

Studer, E.; Coulon, N.; Stietel, A.; Damian, F.; Golfier, H.; Raepsaet, X.

2003-01-01

The CEA R and D program on advanced Gas Cooled Reactors (GCR) relies on different concepts: modular High Temperature Reactor (HTR), its evolution dedicated to hydrogen production (Very High Temperature Reactor) and Gas Cooled Fast Reactors (GCFR). Some key safety questions are related to decay heat removal during potential accident. This is strongly connected to passive natural convection (including gas injection of Helium, CO 2 , Nitrogen or Argon) or forced convection using active safety systems (gas blowers, heat exchangers). To support this effort, thermal-hydraulics computer codes will be necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Accurate and efficient modeling of heat transfer by conduction, convection or thermal radiation as well as energy storage are necessary requirements to obtain a high level of confidence in the thermal-hydraulic simulations. To achieve that goal a thorough validation process has to ve conducted. CEA's CAST3M code dedicated to GCR thermal-hydraulics has been validated against different test cases: academic interaction between natural convection and thermal radiation, small scale in-house THERCE experiments and large scale High Temperature Test Reactor benchmarks such as HTTR-VC benchmark. Coupling with neutronics is also an important modeling aspect for the determination of neutronic parameters such as neutronic coefficient (Doppler, moderator,...), critical position of control rods...CEA's CAST3M and CRONOS2 computer codes allow this coupling and a first example of coupled thermal-hydraulics/neutronics calculations has been performed. Comparison with experimental data will be the next step with High Temperature Test Reactor experimental results at nominal power

Thermal barrier confinement experiments in TMX-U tandem mirror. Revision 1

International Nuclear Information System (INIS)

Simonen, T.C.; Allen, S.L.; Baldwin, D.E.

1984-01-01

In our recent experiments on the TMX-U thermal-barrier device, we achieved the end plugging of axial ion losses up to a central cell density of n/sub c/ = 6 x 10 12 cm -3 . During lower density experiments, we measured the axial potential profile characteristic of a thermal barrier and found an ion-confining potential greater than 1.5 kV and a potential depression of 0.45 kV in the barrier region. The average beta of hot end plug electrons has reached 15% and of hot central cell ions has reached 6%. In addition, we heated deuterium ions in the central cell with ICRF to an average perpendicular energy of 2 keV. During strong end plugging at low density (7 x 10 11 cm -3 ), the axial ion confinement time tau/sub parallel to/ reached 50 to 100 ms while the nonambiopolar radial ion confinement time tau/sub perpendicular to/ was 14 ms - independent of end plugging. Electrically floating end walls doubled the radial ion confinement time. At higher densities and lower potentials, tau/sub parallel to/ was 6 to 12 ms and tau/sub perpendicular to/ exceeded 100 ms

Experience on environmental qualification of safety-related components for Darlington Nuclear Generating Station

International Nuclear Information System (INIS)

Yu, A.S.; Kukreti, B.M.

1987-01-01

The proliferation of Nuclear Power Plant safety concerns has lead to increasing attention over the Environmental Qualification (EQ) of Nuclear Power Plant Safety-Related Components to provide the assurance that the safety related equipment will meet their intended functions during normal operation and postulated accident conditions. The environmental qualification of these components is also a Licensing requirement for Darlington Nuclear Generating Station. This paper provides an overview of EQ and the experience of a pilot project, in the qualification of the Main Moderator System safety-related functions for the Darlington Nuclear Generating Station currently under construction. It addresses the various phases of qualification from the identification of the EQ Safety-Related Components List, definition of location specific service conditions (normal, adbnormal and accident), safety-related functions, Environmental Qualification Assessments and finally, an EQ system summary report for the Main Moderator System. The results of the pilot project are discussed and the methodology reviewed. The paper concludes that the EQ Program developed for Darlington Nuclear Generating Station, as applied to the qualification of the Main Moderator System, contained all the elements necessary in the qualification of safety-related equipment. The approach taken in the qualification of the Moderator safety-related equipment proves to provide a sound framework for the qualification of other safety-related components in the station

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

Energy Technology Data Exchange (ETDEWEB)

Haihua Zhao; Per F. Peterson

2010-10-01

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

The CERN Detector Safety System for the LHC Experiments

CERN Document Server

Lüders, S; Morpurgo, G; Schmeling, S

2003-01-01

The Detector Safety System (DSS), currently being developed at CERN under the auspices of the Joint Controls Project (JCOP), will be responsible for assuring the protection of equipment for the four LHC experiments. Thus, the DSS will require a high degree of both availability and reliability. After evaluation of various possible solutions, a prototype is being built based on a redundant Siemens PLC front-end, to which the safety-critical part of the DSS task is delegated. This is then supervised by a PVSS SCADA system via an OPC server. The PLC front-end is capable of running autonomously and of automatically taking predefined protective actions whenever required. The supervisory layer provides the operator with a status display and with limited online reconfiguration capabilities. Configuration of the code running in the PLCs will be completely data driven via the contents of a "Configuration Database". Thus, the DSS can easily adapt to the different and constantly evolving requirements of the LHC experimen...

Safety concerns for superconducting magnets of upcoming fusion experiments

International Nuclear Information System (INIS)

Turner, L.R.

1983-01-01

-Several fusion experiments being constructed (Tore Supra) or contemplated (DCT 8, Alcator DCT) feature superconducting coils. These coils introduce the following safety concerns: 1. Internally Cooled Conductor (ICC). ICC's are found to be highly stable against short heat pulses, even when the coolant is stagnant or moving at low steady-state velocity. However, a large heat pulse is certain to quench the conductor. Thus, determining the stability limits is vital. 2. Helium II Cooling. Helium II has both unique advantages as a coolant and unique safety problems. 3. Shorted Turns. In magnets with shorts from operational accidents, the current can switch back and forth between the short and the shorted turns, as those alternatively go normal and superconducting. 4. Hybrid Superconducting-Normal Conducting Coil System. The possibility of unequal currents in the different magnets and thus of unexpected forces on the superconducting magnets is much greater than for an all-superconducting system. Analysis of these problems are presented

Physics Based Electrolytic Capacitor Degradation Models for Prognostic Studies under Thermal Overstress

Science.gov (United States)

Kulkarni, Chetan S.; Celaya, Jose R.; Goebel, Kai; Biswas, Gautam

2012-01-01

Electrolytic capacitors are used in several applications ranging from power supplies on safety critical avionics equipment to power drivers for electro-mechanical actuators. This makes them good candidates for prognostics and health management research. Prognostics provides a way to assess remaining useful life of components or systems based on their current state of health and their anticipated future use and operational conditions. Past experiences show that capacitors tend to degrade and fail faster under high electrical and thermal stress conditions that they are often subjected to during operations. In this work, we study the effects of accelerated aging due to thermal stress on different sets of capacitors under different conditions. Our focus is on deriving first principles degradation models for thermal stress conditions. Data collected from simultaneous experiments are used to validate the desired models. Our overall goal is to derive accurate models of capacitor degradation, and use them to predict performance changes in DC-DC converters.

Current MitraClip experience, safety and feasibility in the Netherlands

OpenAIRE

Rahhab, Z.; Kortlandt, F.A.; Velu, J.F.; Schurer, R.A.J.; Delgado, V.; Tonino, P.; Boven, A.J. van; Branden, B.J.L. Van den; Kraaijeveld, A.O.; Voskuil, M.; Hoorntje, J.; Wely, M.H. van; Houwelingen, K. van; Bleeker, G.B.; Rensing, B.

2017-01-01

PURPOSE: Data on MitraClip procedural safety and efficacy in the Netherlands are scarce. We aim to provide an overview of the Dutch MitraClip experience. METHODS: We pooled anonymised demographic and procedural data of 1151 consecutive MitraClip patients, from 13 Dutch hospitals. Data was collected by product specialists in collaboration with local operators. Effect on mitral regurgitation was intra-procedurally assessed by transoesophageal echocardiography. Technical success and device succe...

GOTHIC code simulation of thermal stratification in POOLEX facility

International Nuclear Information System (INIS)

Li, H.; Kudinov, P.

2009-07-01

Pressure suppression pool is an important element of BWR containment. It serves as a heat sink and steam condenser to prevent containment pressure buildup during loss of coolant accident or safety relief valve opening during normal operations of a BWR. Insufficient mixing in the pool, in case of low mass flow rate of steam, can cause development of thermal stratification and reduction of pressure suppression pool capacity. For reliable prediction of mixing and stratification phenomena validation of simulation tools has to be performed. Data produced in POOLEX/PPOOLEX facility at Lappeenranta University of Technology about development of thermal stratification in a large scale model of a pressure suppression pool is used for GOTHIC lumped and distributed parameter validation. Sensitivity of GOTHIC solution to different boundary conditions and grid convergence study for 2D simulations of POOLEX STB-20 experiment are performed in the present study. CFD simulation was carried out with FLUENT code in order to get additional insights into physics of stratification phenomena. In order to support development of experimental procedures for new tests in the PPOOLEX facility lumped parameter pre-test GOTHIC simulations were performed. Simulations show that drywell and wetwell pressures can be kept within safety margins during a long transient necessary for development of thermal stratification. (au)

GOTHIC code simulation of thermal stratification in POOLEX facility

Energy Technology Data Exchange (ETDEWEB)

Li, H.; Kudinov, P. (Royal Institute of Technology (KTH) (Sweden))

2009-07-15

Pressure suppression pool is an important element of BWR containment. It serves as a heat sink and steam condenser to prevent containment pressure buildup during loss of coolant accident or safety relief valve opening during normal operations of a BWR. Insufficient mixing in the pool, in case of low mass flow rate of steam, can cause development of thermal stratification and reduction of pressure suppression pool capacity. For reliable prediction of mixing and stratification phenomena validation of simulation tools has to be performed. Data produced in POOLEX/PPOOLEX facility at Lappeenranta University of Technology about development of thermal stratification in a large scale model of a pressure suppression pool is used for GOTHIC lumped and distributed parameter validation. Sensitivity of GOTHIC solution to different boundary conditions and grid convergence study for 2D simulations of POOLEX STB-20 experiment are performed in the present study. CFD simulation was carried out with FLUENT code in order to get additional insights into physics of stratification phenomena. In order to support development of experimental procedures for new tests in the PPOOLEX facility lumped parameter pre-test GOTHIC simulations were performed. Simulations show that drywell and wetwell pressures can be kept within safety margins during a long transient necessary for development of thermal stratification. (au)

Study on safety of a nuclear ship having an integral marine water reactor. Intelligent information database program concerned with thermal-hydraulic characteristics

International Nuclear Information System (INIS)

Inasaka, Fujio; Nariai, Hideki; Kobayashi, Michiyuki; Murata, Hiroyuki; Aya, Izuo

2001-01-01

As a high economical marine reactor with sufficient safety functions, an integrated type marine water reactor has been considered most promising. At the National Maritime Research Institute, a series of the experimental studies on the thermal-hydraulic characteristics of an integrated/passive-safety type marine water reactor such as the flow boiling of a helical-coil type steam generator, natural circulation of primary water under a ship rolling motion and flashing-condensation oscillation phenomena in pool water has been conducted. This current study aims at making use of the safety analysis or evaluation of a future marine water reactor by developing an intelligent information database program concerned with the thermal-hydraulic characteristics of an integral/passive-safety reactor on the basis of the above-mentioned valuable experimental knowledge. Since the program was created as a Windows application using the Visual Basic, it is available to the public and can be easily installed in the operating system. Main functions of the program are as follows: (1) steady state flow boiling analysis and determination of stability limit for any helical-coil type once-through steam generator design. (2) analysis and comparison with the flow boiling data, (3) reference and graphic display of the experimental data, (4) indication of the knowledge information such as analysis method and results of the study. The program will be useful for the design of not only the future integrated type marine water reactor but also the small sized water reactor. (author)

Experience in the development and practical use of working control levels for radiation safety

International Nuclear Information System (INIS)

Epishin, A.V.

1981-01-01

The experience of development and practical use of working control levels (WCL) of radiation safety in the Gorky region, is discussed. WCL are introduced by ''Radiation Safety Guides'' (RSG-76) and have great practical importance. Regional control levels of radiation safety are determined for certain types of operations implying radioactive hazard and differentiated according to the types of sources applied and types of operation. Dose rates, radioactive contamination of operating surfaces, skin, air and waste water are subject to normalization. Limits of individual radiation doses specified according to operation categories are included. 10 tables of regional WCL indices are developed [ru

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)

System Design Description PFP Thermal Stabilization

International Nuclear Information System (INIS)

RISENMAY, H.R.

2000-01-01

The purpose of this document is to provide a system design description (SDD) and design basis for the Plutonium Finishing Plant (PFP) Thermal Stabilization project. The chief objective of the SDD is to document the Structures, Systems, and Components (SSCs) that establish and maintain the facility Safety Envelope necessary for normal safe operation of the facility; as identified in the FSAR, the OSRs, and Safety Assessment Documents (SADs). This safety equipment documentation should satisfy guidelines for the SDD given in WHC-SD-CP-TI-18 1, Criteria for Identification and Control of Equipment Necessary for Preservation of the Safety Envelope and Safe Operation of PFP. The basis for operational, alarm response, maintenance, and surveillance procedures are also identified and justified in this document. This document and its appendices address the following elements of the PFP Thermal Stabilization project: Functional and design requirements; Design description; Safety Envelope Analysis; Safety Equipment Class; and Operational, maintenance and surveillance procedures

Fusion integral experiments and analysis and the determination of design safety factors - I: Methodology

International Nuclear Information System (INIS)

Youssef, M.Z.; Kumar, A.; Abdou, M.A.; Oyama, Y.; Maekawa, H.

1995-01-01

The role of the neutronics experimentation and analysis in fusion neutronics research and development programs is discussed. A new methodology was developed to arrive at estimates to design safety factors based on the experimental and analytical results from design-oriented integral experiments. In this methodology, and for a particular nuclear response, R, a normalized density function (NDF) is constructed from the prediction uncertainties, and their associated standard deviations, as found in the various integral experiments where that response, R, is measured. Important statistical parameters are derived from the NDF, such as the global mean prediction uncertainty, and the possible spread around it. The method of deriving safety factors from many possible NDFs based on various calculational and measuring methods (among other variants) is also described. Associated with each safety factor is a confidence level, designers may choose to have, that the calculated response, R, will not exceed (or will not fall below) the actual measured value. An illustrative example is given on how to construct the NDFs. The methodology is applied in two areas, namely the line-integrated tritium production rate and bulk shielding integral experiments. Conditions under which these factors could be derived and the validity of the method are discussed. 72 refs., 17 figs., 4 tabs

The Belgian experience on the backfitting and safety upgrading of old operating nuclear power plants

International Nuclear Information System (INIS)

Brognon, T.

1993-01-01

The paper describes the methodology for backfitting and safety upgrading during the reevaluation of the Belgian NPP's: first generation (Doel-1, Doel-2, Tihange-1) and second generation plants (Doel-3, Doel-4, Tihange-2 and Tihange-3). A list of essential safety subjects and topics is given. The experience has proved the feasibility of a safety upgrading of operating NPP without injury to its availability, the benefit of a close cooperation between owner, engineering company and safety authorities throughout the project. A global approach to solving numerous specific deficiencies along with the optimization of the investments regarding the safety improvement of the NPP is suggested. Further increase of the know-how will be achieved through the present Belgian programme along with similar activities abroad. (R.I.)

Numerical modelling of levee stability based on coupled mechanical, thermal and hydrogeological processes

Directory of Open Access Journals (Sweden)

Dwornik Maciej

2016-01-01

Full Text Available The numerical modelling of coupled mechanical, thermal and hydrogeological processes for a soil levee is presented in the paper. The modelling was performed for a real levee that was built in Poland as a part of the ISMOP project. Only four parameters were changed to build different flood waves: the water level, period of water increase, period of water decrease, and period of low water level after the experiment. Results of numerical modelling shows that it is possible and advisable to calculate simultaneously changes of thermal and hydro-mechanical fields. The presented results show that it is also possible to use thermal sensors in place of more expensive pore pressure sensors, with some limitations. The results of stability analysis show that the levee is less stable when the water level decreases, after which factor of safety decreases significantly. For all flooding wave parameters described in the paper, the levee is very stable and factor of safety variations for any particular stage were not very large.

Thermal shock experiment analysis, the use of crack arrest toughness measurements

International Nuclear Information System (INIS)

Miannay, D.; Pellissier-Tanon, A.; Chavaillard, J.P.

1984-06-01

The main purpose of thermal shock experiment is to assess the procedure codified in the ASME XI appendix 1 or RCC-M-B appendix ZG, and allow comparisons with numerical simulations. The analysis of the integrity of the PWR vessel belt line under accidental transients is based on reference curves. The test-piece is a cylinder of SA 508 cl.3 steel. Arrest toughness measured agrees with reference curve

Final report of the 'Nordic thermal-hydraulic and safety network (NOTNET)' - Project

International Nuclear Information System (INIS)

Tuunanen, J.; Tuomainen, M.

2005-04-01

A Nordic network for thermal-hydraulics and nuclear safety research was started. The idea of the network is to combine the resources of different research teams in order to carry out more ambitious and extensive research programs than would be possible for the individual teams. From the very beginning, the end users of the research results have been integrated to the network. Aim of the network is to benefit the partners involved in nuclear energy in the Nordic Countries (power companies, reactor vendors, safety regulators, research units). First task within the project was to describe the resources (personnel, know-how, simulation tools, test facilities) of the various teams. Next step was to discuss with the end users about their research needs. Based on these steps, few most important research topics with defined goals were selected, and coarse road maps were prepared for reaching the targets. These road maps will be used as a starting point for planning the actual research projects in the future. The organisation and work plan for the network were established. National coordinators were appointed, as well as contact persons in each participating organisation, whether research unit or end user. This organisation scheme is valid for the short-term operation of NOTNET when only Nordic organisations take part in the work. Later on, it is possible to enlarge the network e.g. within EC framework programme. The network can now start preparing project proposals and searching funding for the first common research projects. (au)

EXPERIENCE NETWORKING UNIVERSITY OF EDUCATION TRAINING MASTERS SAFETY OF LIFE

OpenAIRE

Elvira Mikhailovna Rebko

2016-01-01

The article discloses experience networking of universities (Herzen State Pedagogical University and Sakhalin State University) in the development and implementation of joint training programs for master’s education in the field of life safety «Social security in the urban environment». The novelty of the work is to create a schematic design of basic educational training program for master’s education in the mode of networking, and to identify effective instructional techniques and conditions...

Experiences of power-up rating in Taiwan

International Nuclear Information System (INIS)

Liao, C. C.; Lai, S. Y.; Chen, Y. B.

2010-10-01

Taiwan has six nuclear power reactors in operation, and two advanced reactors under construction. Measurement Uncertainty Recapture (MU R) type power up rates have been implemented for all the operating units. MU Rs are less than 2 percent and are achieved by using more advanced feedwater flow measurement devices to more precisely measure feedwater flow, which is used to calculate reactor thermal power. The other two types of power up rates are stretch power up rates (SPU) and extended power up rates (EPU). SPU are typically up to 7 percent and EPU are greater that 7 percent but less than 20%. The Atomic Energy Council (Aec) regulates the maximum thermal power level at which a nuclear power plant may operate. In order to increase the rated thermal power of a plant, utility needs to submit an application to the Aec for approval. Detailed safety analysis is required and will be thoroughly reviewed by Aec special task force to ensure the plant safety after implementing the power up rate. Important findings will be documented in the safety evaluation reports. In 2006, Tai power submitted Kuosheng nuclear power plants MU R application, which was the first power up rate application in Taiwan. Till middle of 2009, Tai power has completed the MU R project for all the existing units. The actual thermal power up rates of the six units are whit in the range of 0.3% to 1.5%, resulting in net 56.3 M We increase. Following the success of MU R, Tai power has lunched another project for SPU. In order to enhance the regulatory review process, Aec has drafted a guideline for SPU and EPU by mainly referencing U.S. experience. This guideline shall be beneficial to both licensee and regulatory body in either document preparation or safety review work for the future power up rate applications. (Author)

State of the art on the heat transfer experiments under supercritical pressure condition

International Nuclear Information System (INIS)

Kim, Hwan Yeol; Song, Chul Hwa

2003-07-01

The SCWR(Super-Critical Water cooled Reactor) is one of the six reactor candidates selected in the Gen-IV project which aims at the development of new reactors with enhanced economy and safety. The SCWR is considered to be a feasible concept of new nuclear power plant if the existing technologies developed in fossil fuel fired plant and LWR technologies together with additional research on several disciplines such as materials, water chemistry and safety. As KAERI takes part in the GIF(Generation IV Forum) for the Gen-IV project, domestic concerns about the SCWR have been recently increased. In order to establish a foundation for the development of SCWR, efforts should be concentrated on the conceptual design of systems and the associated key experiments as well. Heat transfer experiments, among others, under supercritical condition are required for the proper prediction of thermal hydraulic phenomena, which are essential for the thermal hydraulic designs of reactor core. Nevertheless, the experiments have not been performed in Korea yet. This report deals with fundamental surveys on the heat transfer experiments under supercritical conditions, which are required for the understanding of heat transfer characteristics for the thermal hydraulic designs of supercritical reactor core. Investigations on the physical properties of water and CO 2 showed that the physical properties such as density, specific heat, viscosity and thermal conductivity are significantly changed near the pseudo-critical points. The state of the art on the heat transfer characteristics in relation with heat transfer deterioration and heat transfer coefficient is briefly described. In addition, previous experiments with supercritical water as well as supercritical CO 2 and Freon used for an alternating fluid are presented

State of the art on the heat transfer experiments under supercritical pressure condition

Energy Technology Data Exchange (ETDEWEB)

Kim, Hwan Yeol; Song, Chul Hwa

2003-07-01

The SCWR(Super-Critical Water cooled Reactor) is one of the six reactor candidates selected in the Gen-IV project which aims at the development of new reactors with enhanced economy and safety. The SCWR is considered to be a feasible concept of new nuclear power plant if the existing technologies developed in fossil fuel fired plant and LWR technologies together with additional research on several disciplines such as materials, water chemistry and safety. As KAERI takes part in the GIF(Generation IV Forum) for the Gen-IV project, domestic concerns about the SCWR have been recently increased. In order to establish a foundation for the development of SCWR, efforts should be concentrated on the conceptual design of systems and the associated key experiments as well. Heat transfer experiments, among others, under supercritical condition are required for the proper prediction of thermal hydraulic phenomena, which are essential for the thermal hydraulic designs of reactor core. Nevertheless, the experiments have not been performed in Korea yet. This report deals with fundamental surveys on the heat transfer experiments under supercritical conditions, which are required for the understanding of heat transfer characteristics for the thermal hydraulic designs of supercritical reactor core. Investigations on the physical properties of water and CO{sub 2} showed that the physical properties such as density, specific heat, viscosity and thermal conductivity are significantly changed near the pseudo-critical points. The state of the art on the heat transfer characteristics in relation with heat transfer deterioration and heat transfer coefficient is briefly described. In addition, previous experiments with supercritical water as well as supercritical CO{sub 2} and Freon used for an alternating fluid are presented.

Experiment and prediction on thermal conductivity of Al2O3/ZnO ...

Indian Academy of Sciences (India)

Administrator

Experiment and prediction on thermal conductivity of Al2O3/ZnO nano thin film interface structure. PING YANG*, LIQIANG ZHANG, HAIYING YANG†, DONGJING LIU and XIALONG LI. Laboratory of Advanced Manufacturing and Reliability for MEMS/NEMS/OEDS,. School of Mechanical Engineering, Jiangsu University, ...

Development of Thermal Radiation Experiments Kit Based on Data Logger for Physics Learning Media

Science.gov (United States)

Permana, H.; Iswanto, B. H.

2018-04-01

Thermal Radiation Experiments Kit (TREK) based on data logger for physics learning media was developed. TREK will be used as a learning medium on the subject of Temperature and Heat to explain the concept of emissivity of a material in grade XI so that it can add variations of experiments which are commonly done such as thermal expansion, transfer of thermal energy (conduction, convection, and radiation), and specific heat capacity. DHT11 sensor is used to measure temperature and microcontroller Arduino-uno used as data logger. The object tested are in the form of coated glass thin films and aluminum with different colors. TREK comes with a user manual and student worksheet (LKS) to make it easier for teachers and students to use. TREK was developed using the ADDIE Development Model (Analyze, Design, Development, Implementation, and Evaluation). And validated by experts, physics teachers, and students. Validation instrument is a questionnaire with a five-item Likert response scale with reviewed aspect coverage: appropriate content and concepts, design, and user friendly. The results showed that TREK was excellent (experts 88.13%, science teachers 95.68%, and students 85.77%).

Fast thermal transients on valve

International Nuclear Information System (INIS)

Ferjancic, M.; Stok, B.; Halilovic, M.; Koc, P.; Mole, N.; Otrin, Z.; Kotar, A.

2007-01-01

One of the regulatory body methods to supervise nuclear safety of a nuclear power plant is a review of plant modifications and evaluation of their impact on plant operating experience. The Slovenian Nuclear Safety Administration (SNSA) licensed in April 2003 the use of leak-before-break (LBB) methodology in the Krsko NPP for the primary loop including surge line and connecting pipelines with minimal diameter of 6 inch. The SNSA decision based also on fracture mechanics analyses that include direct pipe failure mechanisms such as water hammer, creep damage, erosion and corrosion, fatigue and environmental conditions over the entire life of the plant. The evaluation of the operating transients pointed out, that presumed loadings, used for the LBB analysis, did not incorporate all the fast thermal transients data. For that purpose the SNSA requested Faculty of Mechanical Engineering (FS) in Ljubljana to perform additional analyses. The results of the analysis shall confirm the validity of the LBB analysis. (author)

Comprehensive safety analysis code system for nuclear fusion reactors II: Thermal analysis during plasma disruptions for international thermonuclear experimental reactor

International Nuclear Information System (INIS)

Honda, T.; Maki, K.; Okazaki, T.

1994-01-01

Thermal characteristics of a fusion reactor [International Thermonuclear Experimental Reactor (ITER) Conceptual Design Activity] during plasma disruptions have been analyzed by using a comprehensive safety analysis code for nuclear fusion reactors. The erosion depth due to disruptions for the armor of the first wall depends on the current quench time of disruptions occurring in normal operation. If it is possible to extend the time up to ∼50 ms, the erosion depth is considerably reduced. On the other hand, the erosion depth of the divertor is ∼570 μm for only one disruption, which is determined only by the thermal flux during the thermal quench. This means that the divertor plate should be exchanged after about nine disruptions. Counter-measures are necessary for the divertor to relieve disruption influences. As other scenarios of disruptions, beta-limit disruptions and vertical displacement events were also investigated quantitatively. 13 refs., 5 figs

Safety analysis report for the Hanford Critical Mass Laboratory: Supplement No. 2. Experiments with heterogeneous assemblies

International Nuclear Information System (INIS)

Gore, B.F.; Davenport, L.C.

1981-04-01

Factors affecting the safety of criticality experiments using heterogeneous assemblies are described and assessed. It is concluded that there is no substantial change in safety from experiments already being routinely performed at the Critical Mass Laboratory (CML), and that laboratory and personnel safety are adequately provided by the combination of engineered and administrative safety limits enforced at the CML. This conclusion is based on the analysis of operational controls, potential hazards, and the consequences of accidents. Contingencies considered that could affect nuclear criticality include manual changes in fuel loadings, water flooding, fire, explosion, loss of services, earthquake, windstorm, and flood. Other potential hazards considered include radiation exposure to personnel, and potential releases within the Assembly Room and outside to the environment. It is concluded that the Maximum Credible Nuclear Burst of 3 x 10 18 fissions (which served as the design basis for the CML) is valid for heterogeneous assemblies as well as homogeneous assemblies. This is based upon examination of the results of reactor destructive tests and the results of the SL-1 reactor destructive accident. The production of blast effects which might jeopardize the CML critical assembly room (of thick reinforced concrete) is not considered credible due to the extreme circumstances required to produce blast effects in reactor destructive tests. Consequently, it is concluded that, for experiments with heterogeneous assemblies, the consequences of the Maximum Credible Burst are unchanged from those previously estimated for experiments with homogeneous systems

Application of large-eddy simulation to pressurized thermal shock: Assessment of the accuracy

International Nuclear Information System (INIS)