WorldWideScience

Sample records for safety design features

  1. Safety design features of the IRIS

    International Nuclear Information System (INIS)

    2009-01-01

    The International Reactor Innovative and Secure (IRIS) is an advanced, integral, light water cooled reactor of medium generating capacity (335 MW(e)), that features an integral reactor vessel containing all the reactor primary system components, including steam generators, coolant pumps, pressurizer and heaters, and control rod drive mechanisms; in addition to the typical core, internals, control rods and neutron reflector. This integral configuration allows for the use of a small, high design pressure, spherical steel containment which results in a significant reduction in the size of the nuclear island. Other IRIS innovations include a simplified passive safety system concept and equipment features that derive from the 'safety-by-design' philosophy. This design approach allows for elimination of certain accident initiators at the design stage, or when outright elimination is not possible, decreases accident consequences and/or their probability of occurrence. Major design characteristics of the IRIS are given. As part of the IRIS pre-application licensing review by the U.S. Nuclear Regulatory Commission (NRC), the IRIS design team has developed a test plan that will provide the necessary data for safety analysis computer model verification, as well as for verifying the manufacturing feasibility, operability, and durability of new component designs

  2. Safety analysis for key design features of KALIMER-600 design concept

    International Nuclear Information System (INIS)

    Lee, Yong-Bum; Kwon, Y. M.; Kim, E. K.; Suk, S. D.; Chang, W. P.; Joeng, H. Y.; Ha, K. S.; Heo, S.

    2005-03-01

    KAERI is developing the conceptual design of a Liquid Metal Reactor, KALIMER-600 (Korea Advanced LIquid MEtal Reactor) under the Long-term Nuclear R and D Program. KALIMER-600 addresses key issues regarding future nuclear power plants such as plant safety, economics, proliferation, and waste. In this report, key safety design features are described and safety analyses results for typical ATWS accidents, containment design basis accidents, and flow blockages in the KALIMER design are presented. First, the basic approach to achieve the safety goal and main design features of KALIMER-600 are introduced in Chapter 1, and the event categorization and acceptance criteria for the KALIMER-600 safety analysis are described in Chapter 2, In Chapter 3, results of inherent safety evaluations for the KALIMER-600 conceptual design are presented. The KALIMER-600 core and plant system are designed to assure benign performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram (ATWS) have been performed using the SSC-K code to investigate the KALIMER-600 system response to the events. The objectives of Chapter 4, are to assess the response of KALIMER-600 containment to the design basis accidents and to evaluate whether the consequences are acceptable or not in the aspect of structural integrity and the exposure dose rate. In Chapter 5, the analysis of flow blockage for KALIMER-600 with the MATRA-LMR-FB code, which has been developed for the internal flow blockage in a LMR subassembly, are described. The cases with a blockage of 6-subchannel, 24-subchannel, and 54-subchannel are analyzed

  3. The design and safety features of the IRIS reactor

    International Nuclear Information System (INIS)

    Carelli, Mario D.; Conway, L.E.; Oriani, L.; Petrovic, B.; Lombardi, C.V.; Ricotti, M.E.; Barroso, A.C.O.; Collado, J.M.; Cinotti, L.; Todreas, N.E.; Grgic, D.; Moraes, M.M.; Boroughs, R.D.; Ninokata, H.; Ingersoll, D.T.; Oriolo, F.

    2004-01-01

    Salient features of the International Reactor Innovative and Secure (IRIS) are presented here. IRIS, an integral, modular, medium size (335 MWe) PWR, has been under development since the turn of the century by an international consortium led by Westinghouse and including over 20 organizations from nine countries. Described here are the features of the integral design which includes steam generators, pumps and pressurizer inside the vessel, together with the core, control rods, and neutron reflector/shield. A brief summary is provided of the IRIS approach to extended maintenance over a 48-month schedule. The unique IRIS safety-by-design approach is discussed, which, by eliminating accidents, at the design stage, or decreasing their consequences/probabilities when outright elimination is not possible, provides a very powerful first level of defense in depth. The safety-by-design allows a significant reduction and simplification of the passive safety systems, which are presented here, together with an assessment of the IRIS response to transients and postulated accidents

  4. Safety Analysis for Key Design Features of KALIMER-600 Design Concept

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Bum; Kwon, Y. M.; Kim, E. K.; Suk, S. D.; Chang, W. P.; Jeong, H. Y.; Ha, K. S

    2007-02-15

    This report contains the safety analyses of the KALIMER-600 conceptual design which KAERI has been developing under the Long-term Nuclear R and D Program. The analyses have been performed reflecting the design developments during the second year of the 4th design phase in the program. The specific presentations are the key design features with the safety principles for achieving the safety objectives, the event categorization and safety criteria, and results on the safety analyses for the DBAs and ATWS events, the containment performance, and the channel blockages. The safety analyses for both the DBAs and ATWS events have been performed using SSC-K version 1.3., and the results have shown the fulfillment of the safety criteria for DBAs with conservative assumptions. The safety margins as well as the inherent safety also have been confirmed for the ATWS events. For the containment performance analysis, ORIGEN-2.1 and CONTAIN-LMR have been used. In results, the structural integrity has been acceptable and the evaluated exposure dose rate has been complied with 10 CFR 100 and PAG limits. The analysis results for flow blockages of 6-subchannels, 24-subchannels, and 54- subchannels with the MATRA-LMR-FB code, have assured the integrity of subassemblies.

  5. Development of quantitative goals for inherent safety feature design and licensing

    International Nuclear Information System (INIS)

    Kastenberg, W.E.; Apostolakis, G.; Dhir, V.K.; Okrent, D.

    1987-01-01

    There is now considerable interest in the development of advanced fast reactors whose major focus is inherent safety. The achievement of inherent safety can be viewed from several aspects. In the Integral Fast Reactor Concept the approach is to utilize the intrinsic characteristics of pool-type liquid metal fast breeder reactors (LMFBRs) and the properties of metal fuels to integrate a high degree of inherent safety into the design. The PRISM and SAFR concepts focus on other inherent safety features. The reactors discussed above represent a radical departure from existing LWR designs as well as previous LMFBR designs (e.g., CRBRP) which are based, for the most part, on the General Design Criteria found in 10CFR50 Appendix. In view of these parallel developments (advanced reactors exploiting inherent safety and the use of quantitative goals to augment licensing), there appears to be a need to perform research on the development of methods for designing, assessing, and licensing inherent safety features in advanced reactors. The objectives of such research are outlined

  6. The design features and safety concepts of the nuclear heating reactor developed in China

    International Nuclear Information System (INIS)

    Zheng Wenxiang; Wang Dazhong

    1995-01-01

    Based on the specific conditions of the nuclear heat applications and the development objectives of the advanced reactors, the nuclear heating reactor (NHR) exploited in China has adhered to the new safety concepts and been designed with a number of advanced features, including the integrated arrangement, full power natural circulation capacity, self-pressurized performance, dynamically-hydraulic control rod drive and passive safety systems, so that higher standard of safety as well as simplification in the plant systems and improvement in economic viability has been achieved. This paper describes the special consideration in the design as well as the main design features and safety concepts of the NHR. Some experimental and analytical results are also presented to demonstrate the NHR safety features

  7. Preliminary safety analysis for key design features of KALIMER

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, D. H.; Kwon, Y. M.; Chang, W. P.; Suk, S. D.; Lee, S. O.; Lee, Y. B.; Jeong, K. S

    2000-07-01

    KAERI is currently developing the conceptual design of a liquid metal reactor, KALIMER(Korea Advanced Liquid Metal Reactor) under the long-term nuclear R and D program. In this report, descriptions of the KALIMER safety design features and safety analyses results for selected ATWS accidents are presented. First, the basic approach to achieve the safety goal is introduced in chapter 1, and the safety evaluation procedure for the KALIMER design is described in chapter 2. It includes event selection, event categorization, description of design basis events, and beyond design basis events. In chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. The KALIMER core and plant system are designed to assure design performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram(ATWS) have been performed to investigate the KALIMER system response to the events. They are categorized as bounding events(BEs) because of their low probability of occurrence. In chapter 4, the design of the KALIMER containment dome and the results of its performance analysis are presented. The designs of the existing LMR containment and the KALIMER containment dome have been compared in this chapter. Procedure of the containment performance analysis and the analysis results are described along with the accident scenario and source terms. Finally, a simple methodology is introduced to investigate the core kinetics and hydraulic behavior during HCDA in chapter 5. Mathematical formulations have been developed in the framework of the modified bethe-tait method, and scoping analyses have been performed for the KALIMER core behavior during super-prompt critical excursions.

  8. Design of the Control System for Engineered Safety Features of KIJANG Research Reactor

    International Nuclear Information System (INIS)

    Kim, Hagtae; Kim, Jun-Yeon; Chae, Hee-Taek

    2015-01-01

    The purpose of this paper is to design an effective control system for the Engineered Safety Features (ESF) of KJRR such as the Safety Residual Heat Removal System (SRHRS) pumps and Siphon Break Valve (SBV) without an Engineered Safety Features-Component Control System (ESF-CCS). This control system is called a 'local motor starter', because this system controls motors in the SRHRS pumps and SBVs by receiving the signal from Reactor Protection System (RPS) and Alternate Protection System (APS) when the differential pressure or pool level reach the set points. In this paper, the design concepts and requirements of the local motor starter based on the design features of KJRR is proposed. An ESF is a safety system that mitigates consequences of the Anticipated Operational Occurrence (AOO) and Design Basis Accident (DBA). The results of this paper are able to be used for the development of control systems for research reactors similar to KJRR. The precondition for such application is to have a few ESFs and conduct simple logic. The proposed control system called a local motor starter is being designed, and a manufacture of the actual systems is expected in the foreseeable future

  9. Definitions of engineered safety features and related features for nuclear power plants

    International Nuclear Information System (INIS)

    1986-01-01

    In light water moderated, light water cooled nuclear power plants, definitions are given of engineered safety features which are designed to suppress or prevent dispersion of radioactive materials due to damage etc. of fuel at the times of power plant failures, and of related features which are designed to actuate or operate the engineered safety features. Contents are the following: scope of engineered safety features and of related features; classification of engineered safety features (direct systems and indirect systems) and of related features (auxiliaries, emergency power supply, and protective means). (Mori, K.)

  10. An Axiomatic Design Approach of Nanofluid-Engineered Nuclear Safety Features for Generation III+ React

    International Nuclear Information System (INIS)

    Bang, In Cheol; Heo, Gyun Young; Jeong, Yong Hoon; Heo, Sun

    2009-01-01

    A variety of Generation III/III+ reactor designs featuring enhanced safety and improved economics are being proposed by nuclear power industries around the world to solve the future energy supply shortfall. Nanofluid coolants showing an improved thermal performance are being considered as a new key technology to secure nuclear safety and economics. However, it should be noted that there is a lack of comprehensible design works to apply nanofluids to Generation III+ reactor designs. In this work, the review of accident scenarios that consider expected nanofluid mechanisms is carried out to seek detailed application spots. The Axiomatic Design (AD) theory is then applied to systemize the design of nanofluid-engineered nuclear safety systems such as Emergency Core Cooling System (ECCS) and External Reactor Vessel Cooling System (ERVCS). The various couplings between Gen-III/III+ nuclear safety features and nanofluids are investigated and they try to be reduced from the perspective of the AD in terms of prevention/mitigation of severe accidents. This study contributes to the establishment of a standard communication protocol in the design of nanofluid-engineered nuclear safety systems

  11. Safety design

    International Nuclear Information System (INIS)

    Kunitomi, Kazuhiko; Shiozawa, Shusaku

    2004-01-01

    JAERI established the safety design philosophy of the HTTR based on that of current reactors such as LWR in Japan, considering inherent safety features of the HTTR. The strategy of defense in depth was implemented so that the safety engineering functions such as control of reactivity, removal of residual heat and confinement of fission products shall be well performed to ensure safety. However, unlike the LWR, the inherent design features of the high-temperature gas-cooled reactor (HTGR) enables the HTTR meet stringent regulatory criteria without much dependence on active safety systems. On the other hand, the safety in an accident typical to the HTGR such as the depressurization accident initiated by a primary pipe rupture shall be ensured. The safety design philosophy of the HTTR considers these unique features appropriately and is expected to be the basis for future Japanese HTGRs. This paper describes the safety design philosophy and safety evaluation procedure of the HTTR especially focusing on unique considerations to the HTTR. Also, experiences obtained from an HTTR safety review and R and D needs for establishing the safety philosophy for the future HTGRs are reported

  12. Passive Safety Features for Small Modular Reactors

    International Nuclear Information System (INIS)

    Ingersoll, Daniel T.

    2010-01-01

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

  13. Design data and safety features of commerical nuclear power plant

    International Nuclear Information System (INIS)

    Heddleson, F.A.

    1976-06-01

    Design data, safety features, and site characteristics are summarized for 34 nuclear power units in 17 power stations in the United States. Six pages of data are presented for each plant, consisting of thermal-hydraulic and nuclear factors, containment features, emergency-core-cooling systems, site features, circulating water system data, and miscellaneous factors. An aerial perspective is also presented for each plant. This volume covers Light Water Reactors (LWRs) with dockets 50-508 through 50-549, four HTGRs--50-171, 50-267, 50-450/451, 50-463/464, the Atlantic Floating Station 50-477/478, and the Clinch River Breeder 50-537

  14. Safety features of the MAPLE-X10 reactor design

    International Nuclear Information System (INIS)

    Lee, A.G.; Bishop, W.E.; Heeds, W.

    1990-09-01

    The MAPLE-X10 reactor is a D 2 0-reflected, H 2 0-cooled and -moderated pool-type reactor under construction at the Chalk River Nuclear Laboratories. This 10-MW reactor will produce key medical and industrial radio-isotopes such as 99 Mo, 125 I, and 192 Ir. As the prototype for the MAPLE research reactor concept, the reactor incorporates diverse safety features both inherent in the design and in the added engineered systems. The safety requirements are analogous to those of the Canadian CANDU power reactor since standards for the licensing of new research reactors have not been developed yet by the licensing authority in Canada

  15. Safety features of the MAPLE-X10 reactor design

    International Nuclear Information System (INIS)

    Lee, A.G.; Bishop, W.E.; Heeds, W.

    1990-01-01

    This paper reports on the MAPLE-X10 reactor D 2 O-reflected, H 2 O-cooled and -moderated pool- type reactor, under construction at the Chalk River Nuclear Laboratories. This 10-MW will produce key medical and industrial radioisotopes such as 99 Mo, 125 I, and 192 Ir. The prototype for the MAPLE research reactor concept, the reactor incorporates diverse safety features both inherent in the design and in the added engineered systems. The safety requirements are analogous to those of the Canadian CANDU power reactor as standards for the licensing of new research reactors have not been developed by the licensing authority in Canada

  16. Main design and safety features of a 200MW nuclear heating reactor

    International Nuclear Information System (INIS)

    Zheng, Wenxiang; Gao, Zuying; Wang, Dazhong

    1992-01-01

    Inept has been in charge of the development of a nuclear heating reactor since 1980s, which is one of the national key R and D Programs in China. A 5MWt experimental NCR was completed at Inept in 1989 and has operated successfully for space heating since then. In order to realize the commercialization of the NCR, it has been decided to construct a 200MW demonstration NCR in 1993. A number of advanced features, including natural circulation, integrated arrangement, self-pressurized performance, dual vessel structure, hydraulic control rod drive and passive safety systems, have been incorporated into the NCR-200 to achieve its safety goal and economic viability. This makes the NCR safe, simple, reliable, easy-constructed and maintained. At present, the design work of the NCR-200 have shown that its safety characteristics are excellent. The NCR could play an important role in resolving future energy and environmental problems in China. The paper will mainly cover the key design considerations, main technical features and safety analysis results of the NCR-200

  17. A new design concept for offshore nuclear power plants with enhanced safety features

    International Nuclear Information System (INIS)

    Lee, Kihwan; Lee, Kang-Heon; Lee, Jeong Ik; Jeong, Yong Hoon; Lee, Phill-Seung

    2013-01-01

    Highlights: ► A new design concept for offshore nuclear power plants is proposed. ► The total general arrangement for the concept is suggested. ► A new emergency passive containment cooling system (EPCCS) is proposed. ► A new emergency passive reactor-vessel cooling system (EPRVCS) is proposed. ► Safety features against earthquakes, tsunamis, and storms are discussed. - Abstract: In this paper, we present a new concept for offshore nuclear power plants (ONPP) with enhanced safety features. The design concept of a nuclear power plant (NPP) mounted on gravity-based structures (GBSs), which are widely used offshore structures, is proposed first. To demonstrate the feasibility of the concept, a large-scale land-based nuclear power plant model APR1400, which is the most recent NPP model in the Republic of Korea, is mounted on a GBS while minimizing modification to the original features of APR1400. A new total general arrangement (GA) and basic design principles are proposed and can be directly applied to any existing land based large scale NPPs. The proposed concept will enhance the safety of a NPP due to several aspects. A new emergency passive containment cooling system (EPCCS) and emergency passive reactor-vessel cooling system (EPRVCS) are proposed; their features of using seawater as coolant and safety features against earthquakes, Tsunamis, storms, and marine collisions are also described. We believe that the proposed offshore nuclear power plant is more robust than conventional land-based nuclear power plants and it has strong potential to provide great opportunities in nuclear power industries by decoupling the site of construction and that of installation.

  18. Improved safety features in the design of Alto Lazio NPP

    International Nuclear Information System (INIS)

    Bava, G.; Cianciolo, T.; Del Nero, G.

    1988-01-01

    The ALTO LAZIO Nuclear Power Plant, two 1000Mwe units, is a BWR 6/MARK III located about 100 km north of Rome, on the Tyrrhenian Sea Coasts. The construction of the plant started in 1978, but it has recently been stopped by a Government decision following a national referendum, when the units were about 70% completed. This paper is mainly intended to illustrate the major safety features which have been implemented as result of specific requirements issued by the safety authority (ENEA DISP) during the construction permit stage or the subsequent licensing process. One of the tools used to identify the need for design modifications has been a comprehensive reliability analysis of safety system: in the paper the methods used and the major results obtained by this study are briefly presented. Also, the approach used in the investigation of severe accidents and major applications in the area of plant design and emergency procedures are briefly discussed; furthermore the trend toward a simpler mitigation concept is described

  19. Preliminary safety analysis for key design features of KALIMER with breakeven core

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, Do Hee; Kwon, Y. M.; Chang, W. P.; Suk, S. D.; Lee, Y. B.; Jeong, K. S

    2001-06-01

    KAERI is currently developing the conceptual design of a Liquid Metal Reactor, KALIMER (Korea Advanced Liquid MEtal Reactor) under the Long-term Nuclear R and D Program. KALIMER addresses key issues regarding future nuclear power plants such as plant safety, economics, proliferation, and waste. In this report, descriptions of safety design features and safety analyses results for selected ATWS accidents for the breakeven core KALIMER are presented. First, the basic approach to achieve the safety goal is introduced in Chapter 1, and the safety evaluation procedure for the KALIMER design is described in Chapter 2. It includes event selection, event categorization, description of design basis events, and beyond design basis events.In Chapter 3, results of inherent safety evaluations for the KALIMER conceptual design are presented. The KALIMER core and plant system are designed to assure benign performance during a selected set of events without either reactor control or protection system intervention. Safety analyses for the postulated anticipated transient without scram (ATWS) have been performed to investigate the KALIMER system response to the events. In Chapter 4, the design of the KALIMER containment dome and the results of its performance analyses are presented. The design of the existing containment and the KALIMER containment dome are compared in this chapter. Procedure of the containment performance analysis and the analysis results are described along with the accident scenario and source terms. Finally, a simple methodology is introduced to investigate the core energetics behavior during HCDA in Chapter 5. Sensitivity analyses have been performed for the KALIMER core behavior during super-prompt critical excursions, using mathematical formulations developed in the framework of the Modified Bethe-Tait method. Work energy potential was then calculated based on the isentropic fuel expansion model.

  20. ALWR safety approaches and trends. Implementation of passive safety features in the design

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V

    1995-11-01

    Reactor vendors world-wide are examining various advanced light water reactors (ALWR) options to reach utility goals. The amount of information available about each design varies essentially depending on its maturity. Some advanced reactor designs are the evolutionary results of combining old structures, systems and components in new ways, others use innovative solutions. A summary review is given for better understanding of new ALWR design trends and approaches in different countries and subsequent R and D activities. An attempt was made to describe and assess specific innovative and passive features implemented in the leading ALWR designs for further plant design safety improvements. The advantages and disadvantages of these innovations in obtaining reliable systems have been considered. Also, this report indicates the importance of uncertainties remaining and identifies the additional work needed. 51 refs, 27 figs, 7 tabs.

  1. ALWR safety approaches and trends. Implementation of passive safety features in the design

    International Nuclear Information System (INIS)

    Ignatiev, V.

    1995-11-01

    Reactor vendors world-wide are examining various advanced light water reactors (ALWR) options to reach utility goals. The amount of information available about each design varies essentially depending on its maturity. Some advanced reactor designs are the evolutionary results of combining old structures, systems and components in new ways, others use innovative solutions. A summary review is given for better understanding of new ALWR design trends and approaches in different countries and subsequent R and D activities. An attempt was made to describe and assess specific innovative and passive features implemented in the leading ALWR designs for further plant design safety improvements. The advantages and disadvantages of these innovations in obtaining reliable systems have been considered. Also, this report indicates the importance of uncertainties remaining and identifies the additional work needed. 51 refs, 27 figs, 7 tabs

  2. Development of Draft Regulatory Guide on Accident Analysis for Nuclear Power Plants with New Safety Design Features

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Young Seok; Woo, Sweng Woong; Hwang, Tae Suk [KINS, Daejeon (Korea, Republic of); Sim, Suk K; Hwang, Min Jeong [Environment and Energy Technology, Daejeon (Korea, Republic of)

    2016-05-15

    The present paper discusses the development process of the draft version of regulatory guide (DRG) on accident analysis of the NPP having the NSFD and its result. Based on the consideration on the lesson learned from the previous licensing review, a draft regulatory guide (DRG) on accident analysis for NPP with new safety design features (NSDF) was developed. New safety design features (NSDF) have been introduced to the new constructing nuclear power plants (NPP) since the early 2000 and the issuance of construction permit of SKN Units 3 and 4. Typical examples of the new safety features includes Fluidic Device (FD) within Safety Injection Tanks (SIT), Passive Auxiliary Feedwater System (PAFS), ECCS Core Barrel Duct (ECBD) which were adopted in APR1400 design and/or APR+ design to improve the safety margin of the plants for the postulated accidents of interest. Also several studies of new concept of the safety system such as Hybrid ECCS design have been reported. General and/or specific guideline of accident analysis considering the NSDF has been requested. Realistic evaluation of the impact of NSDF on accident with uncertainty and separated accident analysis accounting the NSDF impact were specified in the DRG. Per the developmental process, identification of key issues, demonstration of the DRG with specific accident with specific NSDF, and improvement of DGR for the key issues and their resolution will be conducted.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-31

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

  4. New design of engineered safety features-component control system to improve performance and reliability

    International Nuclear Information System (INIS)

    Kim, S.T.; Jung, H.W.; Lee, S.J.; Cho, C.H.; Kim, D.H.; Kim, H.

    2006-01-01

    Full text: Full text: The Engineered Safety Features-Component Control System (ESF-CCS) controls the engineered safety features of a Nuclear Power Plant such as Solenoid Operated Valves (SOV), Motor Operated Valves (MOV), pumps, dampers, etc. to mitigate the effects of a Design Basis Accident (DBA) or an abnormal operation. ESF-CCS serves as an interface system between the Plant Protection System (PPS) and remote actuation devices. ESF-CCS is composed of fault tolerant Group Controllers GC, Loop Controllers (LC), ESF-CCS Test and Interface Processor (ETIP) and Cabinet Operator Module (COM) and Control Channel Gateway (CCG) etc. GCs in each division are designed to be fully independent triple configuration, which perform system level NSSS and BOP ESFAS logic (2-out-of-4 logic and l-out-of-2 logic, respectively) making it possible to test each GC individually during normal operation. In the existing configuration, the safety-related plant component control is part of the Plant Control System (PCS) non-safety system. For increased safety and reliability, this design change incorporates this part into the LCs, and is therefore designed according to the safety-critical system procedures. The test and diagnosis capabilities of ETIP and COM are reinforced. By means of an automatic periodic test for all main functions of the system, it is possible to quickly determine an abnormal status of the system, and to decrease the elapsed time for tests, thus effectively increasing availability. ESF-CCS consists of four independent divisions (A, B, C, and D) in the Advanced Power Reactor 1400 (APR1400). One prototype division is being manufactured and will be tested

  5. Design and safety features of commercial nuclear power plants in Japan, 1976 edition

    International Nuclear Information System (INIS)

    Izumi, Fumio; Harayama, Yasuo

    1976-10-01

    The December 1975 edition (JAERI-M 5959) contained design particulars and safety features of 20 commercial nuclear power plants in Japan as of December 1974. Subsequently new plants have been put into operation and some plants under construction have undergone design modifications. The present edition presents similar data of the commercial nuclear power plants in Japan up to June 1976, compiled by computer processing. (auth.)

  6. Safety design requirements for safety systems and components of JSFR

    International Nuclear Information System (INIS)

    Kubo, Shigenobu; Shimakawa, Yoshio; Yamano, Hidemasa; Kotake, Shoji

    2011-01-01

    Safety design requirements for JSFR were summarized taking the development targets of the FaCT project and design feature of JSFR into account. The related safety principle and requirements for Monju, CRBRP, PRISM, SPX, LWRs, IAEA standards, goals of GIF, basic principle of INPRO etc. were also taken into account so that the safety design requirements can be a next-generation global standard. The development targets for safety and reliability are set based on those of FaCT, namely, ensuring safety and reliability equal to future LWR and related fuel cycle facilities. In order to achieve these targets, the defence-in-depth concept is used as the basic safety design principle. General features of the safety design requirements are 1) Achievement of higher reliability, 2) Achievement of higher inspectability and maintainability, 3) Introduction of passive safety features, 4) Reduction of operator action needs, 5) Design consideration against Beyond Design Basis Events, 6) In-Vessel Retention of degraded core materials, 7) Prevention and mitigation against sodium chemical reactions, and 8) Design against external events. The current specific requirements for each system and component are summarized taking the basic design concept of JSFR into account, which is an advanced loop-type large-output power plant with a mixed-oxide-fuelled core. (author)

  7. Safety Design Approach for the Development of Safety Requirements for Design of Commercial HTGR

    International Nuclear Information System (INIS)

    Ohashi, Hirofumi; Sato, Hiroyuki; Nakagawa, Shigeaki; Tachibana, Yukio; Nishihara, Tetsuo; Yan, Xing; Sakaba, Nariaki; Kunitomi, Kazuhiko

    2014-01-01

    The research committee on “Safety requirements for HTGR design” was established in 2013 under the Atomic Energy Society of Japan to develop the draft safety requirements for the design of commercial High Temperature Gas-cooled Reactors (HTGRs), which incorporate the HTGR safety features demonstrated using the High Temperature Engineering Test Reactor (HTTR), lessons learned from the accident of Fukushima Daiichi Nuclear Power Station and requirements for the integration of the hydrogen production plants. The safety design approach for the commercial HTGRs which is a basement of the safety requirements is determined prior to the development of the safety requirements. The safety design approaches for the commercial HTGRs are to confine the radioactive materials within the coated fuel particles not only during normal operation but also during accident conditions, and the integrity of the coated fuel particles and other requiring physical barriers are protected by the inherent and passive safety features. This paper describes the main topics of the research committee, the safety design approaches and the safety functions of the commercial HTGRs determined in the research committee. (author)

  8. Integral fast reactor concept inherent safety features

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

    1987-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFT development effort are improved economics and enhanced safety. The design features that together fulfill these goals are: 1) a liquid metal (sodium) coolant, 2) a pool-type reactor primary system configuration, 3) an advanced ternary alloy metallic fuel, and 4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS)

  9. Integral Fast Reactor concept inherent safety features

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

    1986-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. The design features that together fulfill these goals are: (1) a liquid metal (sodium) coolant, (2) a pool-type reactor primary system configuration, (3) an advanced ternary alloy metallic fuel, and (4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS)

  10. EC6 safety design improvements

    Energy Technology Data Exchange (ETDEWEB)

    Yu, S.; Lee, A.G.; Soulard, M. [Candu Energy Inc., Mississauga, ON (Canada)

    2014-07-01

    The Enhanced CANDU 6 (EC6) builds on the proven high performance design such as the Qinshan CANDU 6 reactor, and has made improvements to safety, operational performance, and has incorporated extensive operational feedback. Completion of all three phases of the pre-licensing design review by the Canadian Regulator - the Canadian Nuclear Safety Commission has provided a higher level of assurance that the EC6 reference design has taken modern regulatory requirements and expectations into account and further confirmed that there are no fundamental barriers to licensing the EC6 design in Canada. The EC6 design is based on the defence-in-depth principles in INSAG-10 and provides further safety features that address the lessons learned from Fukushima. With these safety features, the EC6 design has strengthened accident prevention as the first priority in the defence-in-depth strategy, as outlined in INSAG-10. As well, the EC6 design has incorporated further mitigation measures to provide additional protection of the public and the environment if the preventive measures fail. The EC6 design has an appropriate combination of inherent, passive safety characteristics, engineered features and administrative safety measures to effectively prevent and mitigate severe accident progressions. A strong contributor to the robustness and redundancy of CANDU design is the two-group separation philosophy. This ensures a high degree of independence between safety systems as well as physical separation and functional independence in how fundamental safety functions are provided. This paper will describe the following safety features based on the application of defence-in-depth and design approach to prevent beyond design basis events progressing to severe accidents and to mitigate the consequences if it occurs: Improved steam generator heat sink via a more reliable emergency heat removal system; Increased time before manual field actions are required via enhanced capacity of

  11. Accommodation of unprotected accidents by inherent safety design features in metallic and oxide-fueled LMFBRs

    International Nuclear Information System (INIS)

    Su, S.F.; Cahalan, J.E.; Sevy, R.H.

    1985-01-01

    This paper presents the results of a systematic study of the effectiveness of intrinsic design features to mitigate the consequences of unprotected accidents in metallic and oxide-fueled LMFBRs. The accidents analyzed belong to the class generally considered to lead to core disruption; unprotected loss-of-flow (LOF) and transient over-power (TOP). The results of the study demonstrate the potential for design features to meliorate accident consequences, and in some cases to render them benign. Emphasis is placed on the relative performance of metallic and oxide-fueled core designs, and safety margins are quantified in sensitivity studies. All analyses were carried out using the SASSYS LMFBR systems analysis code (1)

  12. Safety design of Qinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Ouyang Yu; Zhang Lian; Du Shenghua; Zhao Jiayu

    1984-01-01

    Safety issues have been greatly emphasized through the design of the Qinshan Nuclear Power Plant. Reasonable safety margine has been taken into account in the plant design parameters, the design incorporated various safeguard systems, such as engineering safety feature systems, safety protection systems and the features to resist natural catastrophes, e. g. earthquake, hurricanes, tide and so on. Preliminary safety analysis and environmental effect assessment have been done and anti-accident provisions and emergency policy were carefully considered. Qinshan Nuclear Power Plant safety related systems are designed in accordance with the common international standards established in the late 70's, as well as the existing engineering standard of China

  13. Engineering safety features for high power experimental reactors

    International Nuclear Information System (INIS)

    Doval, A.; Villarino, E.; Vertullo, A.

    2000-01-01

    In the present analysis we will focus our attention in the way engineering safety features are designed in order to prevent fuel damage in case of abnormal or accidental situations. To prevent fuel damage two main facts must be considered, the shutdown of the reactor and the adequate core cooling capacity, it means that both, neutronic and thermohydraulic aspects must be analysed. Some neutronic safety features are common to all power ranges like negative feedback reactivity coefficients and the required number of control rods containing the proper absorber material to shutdown the reactor. From the thermohydraulic point of view common features are siphon-breaker devices and flap valves for those powers requiring cooling in the forced convection regime. For the high power reactor group, the engineering safety features specially designed for a generic reactor of 20 MW, will be presented here. From the neutronic point of view besides the common features, and to comply with our National Regulatory Authority, a Second Shutdown System was designed as a redundant shutdown system in case the control plates fail. Concerning thermohydraulic aspects besides the pump flywheels and the flap valves providing the natural convection loop, a metallic Chimney and a Chimney Water Injection System were supplied. (author)

  14. Integral fast reactor safety features

    International Nuclear Information System (INIS)

    Cahalan, J.E.; Kramer, J.M.; Marchaterre, J.F.; Mueller, C.J.; Pedersen, D.R.; Sevy, R.H.; Wade, D.C.; Wei, T.Y.C.

    1988-01-01

    The integral fast reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. In addition to liquid metal cooling, the principal design features that distinguish the IFR are: a pool-type primary system, and advanced ternary alloy metallic fuel, and an integral fuel cycle with on-site fuel reprocessing and fabrication. This paper focuses on the technical aspects of the improved safety margins available in the IFR concept. This increased level of safety is made possible by the liquid metal (sodium) coolant and pool-type primary system layout, which together facilitate passive decay heat removal, and a sodium-bonded metallic fuel pin design with thermal and neutronic properties that provide passive core responses which control and mitigate the consequences of reactor accidents

  15. Safety features and licensing of CNNC-ACP100

    Energy Technology Data Exchange (ETDEWEB)

    Zhong, F., E-mail: Zhongfj2000@163.com [Nuclear Power Inst. of China, National Key Lab. of Science and Technology on Reactor System Design Technology (China)

    2014-07-01

    ACP100 is an innovatory modular pressurized water reactor, the engineering safety systems fully adopt passive safety design technology. Its inherent safety and passive features/systems are verified via testing facilities and are highlighted at certain levels of defence in depth. The licensing of ACP 100 is within current LWR framework and meets up-to-date codes and requirements in nuclear safety. (author)

  16. Design Review Report for formal review of safety class features of exhauster system for rotary mode core sampling

    International Nuclear Information System (INIS)

    JANICEK, G.P.

    2000-01-01

    Report documenting Formal Design Review conducted on portable exhausters used to support rotary mode core sampling of Hanford underground radioactive waste tanks with focus on Safety Class design features and control requirements for flammable gas environment operation and air discharge permitting compliance

  17. Design Review Report for formal review of safety class features of exhauster system for rotary mode core sampling

    Energy Technology Data Exchange (ETDEWEB)

    JANICEK, G.P.

    2000-06-08

    Report documenting Formal Design Review conducted on portable exhausters used to support rotary mode core sampling of Hanford underground radioactive waste tanks with focus on Safety Class design features and control requirements for flammable gas environment operation and air discharge permitting compliance.

  18. Integral fast reactor safety features

    International Nuclear Information System (INIS)

    Cahalan, J.E.; Kramer, J.M.; Marchaterre, J.F.; Mueller, C.J.; Pedersen, D.R.; Sevy, R.H.; Wade, D.C.; Wei, T.Y.C.

    1988-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. In addition to liquid metal cooling, the principal design features that distinguish the IFR are: (1) a pool-type primary system, (2) an advanced ternary alloy metallic fuel, and (3) an integral fuel cycle with on-site fuel reprocessing and fabrication. This paper focuses on the technical aspects of the improved safety margins available in the IFR concept. This increased level of safety is made possible by (1) the liquid metal (sodium) coolant and pool-type primary system layout, which together facilitate passive decay heat removal, and (2) a sodium-bonded metallic fuel pin design with thermal and neutronic properties that provide passive core responses which control and mitigate the consequences of reactor accidents

  19. Safety aspects and shield design of a Poton irradiator

    International Nuclear Information System (INIS)

    Mehta, S.K.; Nayak, A.R.; Bongirwar, D.R.; Modi, R.K.; Ramkumar, M.S.

    1998-01-01

    An irradiation plant, POTON, for irradiation of potatoes and onions is being set up at Nashik. Shield design and safety features of this plant incorporate some novel and innovative features like a compact cell, curved cell boundaries for smooth conveyor movement though the cell labyrinth and conform to ICRP and AERB design safety requirements. The safety features include multiple safety interlocks, audio-visual alarms, scram switches and trip wire for avoiding accidental exposures. (author)

  20. Design data and safety features of commercial nuclear power plants. Vol. IV. Dockets 50-452 through 50-503

    International Nuclear Information System (INIS)

    Heddleson, F.A.

    1975-03-01

    Design data, safety features, and site characteristics are summarized for 36 nuclear power units in 18 power stations in the United States. Six pages of data are presented for each plant consisting of thermal-hydraulic and nuclear factors, containment features, emergency-core-cooling systems, site features, circulating water system data, and miscellaneous factors. An aerial perspective is also presented for each plant. The volume covers reactors with dockets 50-452 through 50-503. (U.S.)

  1. What price safety. A probabilistic cost-benefit evaluaton of existing engineered safety features

    International Nuclear Information System (INIS)

    O'Donnell, E.P.

    1978-01-01

    The paper provides a method for performing quantitative cost-benefit evaluations for nuclear safety concerns involving accidents of low probability and potentially large consequences. It presents an application of the method to ECCS, containment, emergency power system and hydrogen recombiner system. This evaluation provides a valuable assessment of the relative cost effectiveness of these features in reducing accident risk. It also provides insight into the sensitivity of cost-benefit calculations to the manner in which safety features are sequantially added in design. (author)

  2. ELFR: The European Lead Fast Reactor. Design, Safety Approach and Safety Characteristics

    International Nuclear Information System (INIS)

    Alemberti, Alessandro

    2012-01-01

    • In the framework of the LEADER project, the safety approach for a Lead cooled fast reactor has been defined and, in particular, all the possible challenges to the main safety functions and their mechanisms have been specified, in order to better define the needed provisions. • On the basis of the above and taking into account the results of the safety analyses performed during previous project (ELSY), a reference configuration of the ELFR plant has been consolidated, by improving and updating the plant design features. In particular, the emerged safety concerns have been analyzed in the LEADER project and a new set of design options and safety provisions have been proposed. • The combination of favourable Lead coolant inherent characteristics and plant design features, specifically developed to face identified challenges, resulted in a very robust and forgiving design, even in very extreme conditions, as a Fukushima-like scenario

  3. Development of ABWR-2 and its safety design

    International Nuclear Information System (INIS)

    Takafumi, Anegawa; Kenji, Tateiwa

    2002-01-01

    This paper reports the current status of development project on ABWR-II, a next generation reactor design based on ABWR, and its safety design. This project was initiated over a decade ago and has completed three phases to date. In Phase I (1991-92), basic design requirements were discussed and several plant concepts were studied. In Phase II (1993-95), key design features were selected in order to establish a reference reactor concept. In Phase III (1996-2000), based on the reference reactor concept, modifications and improvements were made to fulfill the design requirements. By adopting large electric output (1 700 MW), large fuel bundle, modified ECCS, and passive heat removal systems, among other design features, we achieved a design concept capable of increasing both economic competitiveness and safety performance. Main focus of this paper will be on the safety design, safety performance, and further research needs related to safety. (authors)

  4. Design features of SMART for barge mounted application

    International Nuclear Information System (INIS)

    Doo-Jeong Lee; Ju-Hyeon Yoon; Ju-Pyung Kim; Jong-In Kim; Moon-Hee Chang

    2000-01-01

    SMART is an integral reactor of 330 MWt capacity with passive safety features being developed for a wide range of applications including the barge mounted co-generation plant. Its design strives to combine the firmly-established commercial reactor design with new advanced technologies. Thus the use of the industry proven KOFA (Korea Optimized Fuel Assembly) based nuclear fuels is pursued while such radically new technologies as self-pressurizing pressurizer, helical once-through steam generators, and advanced control concepts are being developed. The safety of SMART centers around enhancing the inherent safety characteristics of the reactor and salient features include low core power density, integral arrangement to eliminate large break loss of coolant accident, etc. The progression of emergency situations into accidents is prevented with a number of advanced engineered safety features such as Passive Residual Heat Removal System, Passive Emergency Core Cooling System, Safeguard Vessel, Passive Containment Over-pressure Protection. This paper presents the status of current SMART development, characteristics of SMART safety systems and the possibility of SMART application to barge mounted environment. (author)

  5. Safety performance of preliminary KALIMER conceptual design

    Energy Technology Data Exchange (ETDEWEB)

    Hahn Dohee; Kim Kyoungdoo; Kwon Youngmin; Chang Wonpyo; Suk Soodong [Korea atomic Energy Resarch Inst., Taejon (Korea)

    1999-07-01

    The Korea Atomic Energy Research Institute (KAERI) is developing KALIMER (Korea Advanced Liquid Metal Reactor), which is a sodium cooled, 150 MWe pool-type reactor. The safety design of KALIMER emphasizes accident prevention by using passive processes, which can be accomplished by the safety design objectives including the utilization of inherent safety features. In order to assess the effectiveness of the inherent safety features in achieving the safety design objectives, a preliminary evaluation of ATWS performance for the KALIMER design has been performed with SSC-K code, which is a modified version of SSC-L code. KAERI's modification of the code includes development of reactivity feedback models for the core and a pool model for KALIMER reactor vessel. This paper describes the models for control rod driveline expansion, gas expansion module and the thermal hydraulic model for reactor pool and the results of preliminary analyses for unprotected loss of flow and loss o heat sink. (author)

  6. Safety performance of preliminary KALIMER conceptual design

    International Nuclear Information System (INIS)

    Hahn Dohee; Kim Kyoungdoo; Kwon Youngmin; Chang Wonpyo; Suk Soodong

    1999-01-01

    The Korea Atomic Energy Research Institute (KAERI) is developing KALIMER (Korea Advanced Liquid Metal Reactor), which is a sodium cooled, 150 MWe pool-type reactor. The safety design of KALIMER emphasizes accident prevention by using passive processes, which can be accomplished by the safety design objectives including the utilization of inherent safety features. In order to assess the effectiveness of the inherent safety features in achieving the safety design objectives, a preliminary evaluation of ATWS performance for the KALIMER design has been performed with SSC-K code, which is a modified version of SSC-L code. KAERI's modification of the code includes development of reactivity feedback models for the core and a pool model for KALIMER reactor vessel. This paper describes the models for control rod driveline expansion, gas expansion module and the thermal hydraulic model for reactor pool and the results of preliminary analyses for unprotected loss of flow and loss o heat sink. (author)

  7. Safety assessment in plant layout design using indexing approach: Implementing inherent safety perspective

    International Nuclear Information System (INIS)

    Tugnoli, Alessandro; Khan, Faisal; Amyotte, Paul; Cozzani, Valerio

    2008-01-01

    Layout planning plays a key role in the inherent safety performance of process plants since this design feature controls the possibility of accidental chain-events and the magnitude of possible consequences. A lack of suitable methods to promote the effective implementation of inherent safety in layout design calls for the development of new techniques and methods. In the present paper, a safety assessment approach suitable for layout design in the critical early phase is proposed. The concept of inherent safety is implemented within this safety assessment; the approach is based on an integrated assessment of inherent safety guideword applicability within the constraints typically present in layout design. Application of these guidewords is evaluated along with unit hazards and control devices to quantitatively map the safety performance of different layout options. Moreover, the economic aspects related to safety and inherent safety are evaluated by the method. Specific sub-indices are developed within the integrated safety assessment system to analyze and quantify the hazard related to domino effects. The proposed approach is quick in application, auditable and shares a common framework applicable in other phases of the design lifecycle (e.g. process design). The present work is divided in two parts: Part 1 (current paper) presents the application of inherent safety guidelines in layout design and the index method for safety assessment; Part 2 (accompanying paper) describes the domino hazard sub-index and demonstrates the proposed approach with a case study, thus evidencing the introduction of inherent safety features in layout design

  8. EC6 design features and pre-project licensing review

    Energy Technology Data Exchange (ETDEWEB)

    Yu, S.; Lee, A.G.; Dinh, N.B.; Soulard, M. [CANDU Energy Inc., Mississauga, Ontario, (Canada)

    2013-07-01

    The Enhanced CANDU 6 (EC6) is the new Generation III CANDU reactor design that meets the most up to date Canadian regulatory requirements and customer expectations. Candu Energy Inc. is finalizing development of the EC6 which incorporates the CANDU 6's well-proven features, and adds enhancements that strengthened reactor safety margin and improved operability. The EC6 builds on the proven high performance design and the defence-in-depth features of CANDU 6 units, and has incorporated extensive operational feedback including lessons learned from Fukushima. This paper will provide status of the engineering program including progress on the pre-licensing review of the EC6 design by the Canadian Regulator, CNSC, and will also highlight the design and safety enhancements incorporated in the EC6 product. Safety enhancements to meet safety goals and to improve robustness of systems to respond to design basis accidents and beyond design basis accidents include: new severe accident recovery and heat removal system; improved emergency heat removal system; faster shutoff rods with improved safety margins; mechanical guaranteed shutdown rods; daily load cycling capability; robust containment with containment filter venting system; and improved backed-up electrical supply and cooling services. (author)

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  10. Technical considerations for the development of an engineering safety features control system with PLC

    International Nuclear Information System (INIS)

    Lee, C. K.; Kim, C. H.; Han, J. B.; Kim, H.; Lee, S. S.

    2002-01-01

    Technical considerations are summarized for the development of an ESFCS(Engineered Safety Features Control System) with PLC (Programmable Logic Controller). The ESFCS is required for the mitigation of plant accident conditions and therefore developed in conformance with the design requirements applied to the safety critical system. The design of ESFCS primarily considered its safety, and the system has an architecture that will be able to minimize spurious actuation. The PLC based functional distribution and redundant design features are adopted, and the fieldbus is applied in the communication of information and control signals between PLC processors. It is expected that the ESFCS will have several advanced design features compared with the conventional systems supplied by foreign vendors

  11. The safety designs for the TITAN reversed-field pinch reactor study

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Cheng, E.T.; Creedon, R.L.; Hoot, C.G.; Schultz, K.R.; Grotz, S.P.; Blanchard, J.; Sharafat, S.; Najmabadi, F.

    1989-01-01

    TITAN is a study to investigate the potential of the reversed-field pinch concept as a compact, high-power density energy system. Two reactor concepts were developed, a self-cooled lithium design with vanadium structure and an aqueous solution loop-in-pool design, both operating at 18 MW/m 2 . The key safety features of the TITAN-I lithium-vanadium blanket design are in material selection, fusion power core configuration selection, lithium piping connections, and passive lithium drain tank system. Based on these safety features and results from accident evaluation, TITAN-I can at least be rated at a level 3 of safety assurance. For the TITAN-II aqueous loop-in-pool design, the key passive feature is the complete submersion of the fusion power core and the corresponding primary coolant loop system into a pool of low temperature water. Based on this key safety design feature, the TITAN-II design can be rated at a level 2 of safety assurance. (orig.)

  12. The safety designs for the TITAN reversed-field pinch reactor study

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Cheng, E.T.; Creedon, R.L.; Hoot, C.G.; Schultz, K.R.; Grotz, S.P.; Blanchard, J.P.; Sharafat, S.; Najmabadi, F.

    1988-01-01

    TITAN is a study to investigate the potential of the reversed-field pinch concept as a compact, high-power density energy system. Two reactor concepts were developed, a self-cooled lithium design with vanadium structure and an aqueous solution loop-in-pool design, both operating at 18 MW/m 2 . The key safety features of the TITAN-I lithium-vanadium blanket design are in material selection, fusion power core configuration selection, lithium piping connections and passive lithium drain tank system. Based on these safety features and results from accident evaluation, TITAN-I can at least be rated as level 3 of safety assurance. For the TITAN-II aqueous loop-in-pool design, the key passive feature is the complete submersion of the fusion power core and the corresponding primary coolant loop system into a pool of low temperature water. Based on this key safety design feature, the TITAN-II design can be rated as level 2 of safety assurance. 7 refs., 2 figs

  13. Safety features and research needs of westinghouse advanced reactors

    International Nuclear Information System (INIS)

    Carelli, M.D.; Winters, J.W.; Cummins, W.E.; Bruschi, H.J.

    2002-01-01

    The three Westinghouse advanced reactors - AP600, AP1000 and IRIS - are at different levels of readiness. AP600 has received a Design Certification, its larger size version AP1000 is currently in the design certification process and IRIS has just completed its conceptual design and will initiate soon a licensing pre-application. The safety features of the passive designs AP600/AP1000 are presented, followed by the features of the more revolutionary IRIS, a small size modular integral reactor. A discussion of the IRIS safety by design approach is given. The AP600/AP1000 design certification is backed by completed testing and development which is summarized, together with a research program currently in progress which will extend AP600 severe accident test data to AP1000 conditions. While IRIS will of course rely on applicable AP600/1000 data, a very extensive testing campaign is being planned to address all the unique aspects of its design. Finally, IRIS plans to use a risk-informed approach in its licensing process. (authors)

  14. Inherent safety features in balance-of-plant layout

    International Nuclear Information System (INIS)

    Wattelet, P.L.; Green, K.J.

    1992-01-01

    Future nuclear units must be more economical to construct and operate, and, at the same time, clearly incorporate advances in safety over the current generation of light water reactors. To achieve these goals, the root causes of safety issues must be addressed. In this way, global, cost-effective solutions can be implemented. With simple, direct design approaches, the licensing risk is minimized and configuration control is enhanced. With proper planning in the early stages of plant design, postulated accidents and events can often be mitigated by passive features inherent in the basic structure and layout, eliminating expensive added protective structures and components often found in current designs. Korea Electric Power Corporation's Yonggwang (YGN) Units 3 and 4, shown in an artist's rendering in Figure 1, are now under construction in Korea. Engineering is more than 85% complete, and Unit 3 construction is more than 50% complete. Significant steps toward design simplification and safety enhancement have been made by addressing safety concerns very early in the design effort. The tools used to achieve this were improved symmetry and separation, isolation of potential hazards, and an improved design process

  15. SAFETY BASIS DESIGN DEVELOPMENT CHALLENGES IMECE2007-42747

    Energy Technology Data Exchange (ETDEWEB)

    RYAN GW

    2007-09-24

    'Designing in Safety' is a desired part of the development of any new potentially hazardous system, process, or facility. It is a required part of nuclear safety activities as specified in the U.S. Department of Energy (DOE) Order 420.B, Facility Safety. This order addresses the design of nuclear related facilities developed under federal regulation IOCFR830, Nuclear Safety Management. IOCFR830 requires that safety basis documentation be provided to identify how nuclear safety is being adequately addressed as a condition for system operation (e.g., the safety basis). To support the development of the safety basis, a safety analysis is performed. Although the concept of developing a design that addresses 'Safety is simple, the execution can be complex and challenging. This paper addresses those complexities and challenges for the design activity of a system to treat sludge, a corrosion product of spent nuclear fuel, at DOE's Hanford Site in Washington State. The system being developed is referred to as the Sludge Treatment Project (STP). This paper describes the portion of the safety analysis that addresses the selection of design basis events using the experience gained from the STP and the development of design requirements for safety features associated with those events. Specifically, the paper describes the safety design process and the application of the process for two types of potential design basis accidents associated with the operation of the system, (1) flashing spray leaks and (2) splash and splatter leaks. Also presented are the technical challenges that are being addressed to develop effective safety features to deal with these design basis accidents.

  16. SAFETY BASIS DESIGN DEVELOPMENT CHALLENGES IMECE2007-42747

    International Nuclear Information System (INIS)

    RYAN GW

    2007-01-01

    'Designing in Safety' is a desired part of the development of any new potentially hazardous system, process, or facility. It is a required part of nuclear safety activities as specified in the U.S. Department of Energy (DOE) Order 420.B, Facility Safety. This order addresses the design of nuclear related facilities developed under federal regulation IOCFR830, Nuclear Safety Management. IOCFR830 requires that safety basis documentation be provided to identify how nuclear safety is being adequately addressed as a condition for system operation (e.g., the safety basis). To support the development of the safety basis, a safety analysis is performed. Although the concept of developing a design that addresses 'Safety is simple, the execution can be complex and challenging. This paper addresses those complexities and challenges for the design activity of a system to treat sludge, a corrosion product of spent nuclear fuel, at DOE's Hanford Site in Washington State. The system being developed is referred to as the Sludge Treatment Project (STP). This paper describes the portion of the safety analysis that addresses the selection of design basis events using the experience gained from the STP and the development of design requirements for safety features associated with those events. Specifically, the paper describes the safety design process and the application of the process for two types of potential design basis accidents associated with the operation of the system, (1) flashing spray leaks and (2) splash and splatter leaks. Also presented are the technical challenges that are being addressed to develop effective safety features to deal with these design basis accidents

  17. Conceptual Design of Hybrid Safety Features for NPP by Utilizing Solar Updraft Tower

    Energy Technology Data Exchange (ETDEWEB)

    Song, Sub Lee [Handong Global University, Pohang (Korea, Republic of); Choi, Young Jae; Kim, Yong Jin [KAIST, Daejeon (Korea, Republic of); Park, Hyo Chan; Park, Youn Won [BEES, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, hybrid safety features for NPP with solar updraft tower (SUT) is conceptually suggested to cope with loss of ultimate heat sink accident. The hybrid safety features utilizing SUT target NPPs in seashore of Arabian Gulf. Usually NPPs are constructed near seashore to utilize sea water as an ultimate heat sink. Residual heat or decay heat of nuclear reactor will diffuse into the ocean through the condenser. NPPs in Middle East are expected to be placed in seashore of Arabian Gulf. The NPP site of Barakah is an actual example. For NPPs in seashore of Arabian Gulf, an additional safety concern should be considered. Arabian Gulf is the largest oil transporting route in the world. The oil spill risk in Arabian Gulf will be the largest simultaneously. Unfortunately, not like other oceans, Arabian Gulf is a kind of closed ocean which does not have strong ocean currents connected to out of the gulf. If once oil spill is occurred, its influence can be propagated more than our expectation. The spilled oil also can affect to NPPs in seashore by covering surfaces of condenser. It will directly cause loss of ultimate heat sink. The hybrid safety features of SUT system are expected to aid normal operation of safety system and mitigate consequence of severe accident. Detail analysis and technology development is ongoing now.

  18. Conceptual Design of Hybrid Safety Features for NPP by Utilizing Solar Updraft Tower

    International Nuclear Information System (INIS)

    Song, Sub Lee; Choi, Young Jae; Kim, Yong Jin; Park, Hyo Chan; Park, Youn Won

    2016-01-01

    In this study, hybrid safety features for NPP with solar updraft tower (SUT) is conceptually suggested to cope with loss of ultimate heat sink accident. The hybrid safety features utilizing SUT target NPPs in seashore of Arabian Gulf. Usually NPPs are constructed near seashore to utilize sea water as an ultimate heat sink. Residual heat or decay heat of nuclear reactor will diffuse into the ocean through the condenser. NPPs in Middle East are expected to be placed in seashore of Arabian Gulf. The NPP site of Barakah is an actual example. For NPPs in seashore of Arabian Gulf, an additional safety concern should be considered. Arabian Gulf is the largest oil transporting route in the world. The oil spill risk in Arabian Gulf will be the largest simultaneously. Unfortunately, not like other oceans, Arabian Gulf is a kind of closed ocean which does not have strong ocean currents connected to out of the gulf. If once oil spill is occurred, its influence can be propagated more than our expectation. The spilled oil also can affect to NPPs in seashore by covering surfaces of condenser. It will directly cause loss of ultimate heat sink. The hybrid safety features of SUT system are expected to aid normal operation of safety system and mitigate consequence of severe accident. Detail analysis and technology development is ongoing now

  19. Cost reduction and safety design features of ABWR-II. Annex 5

    International Nuclear Information System (INIS)

    Koh, F.; Moriya, K.; Anegawa, T.

    2002-01-01

    The ABWR-II, which is aimed to be the next generation reactor following the latest BWR: Advanced Boiling Reactor (ABWR), is now under development jointly by the Japanese BWR utilities, General Electric Company, Hitachi Limited, and Toshiba Corporation. The key objectives of ABWR-II development include improvement in economics and further sophistication in safety for commercialization in the late 2010's and after. This paper summarizes the current status of ABWR-II development focusing on economics and safety. Plant power rating, fuel size, CRD rationalization and outage period are discussed from a cost reduction perspective. In terms of safety, the features such as diversification in emergency power sources and passive system application against severe accidents are being introduced. (author)

  20. Passive safety features in current and future water cooled reactors

    International Nuclear Information System (INIS)

    1990-11-01

    Better understanding of the passive safety systems and components in current and future water-cooled reactors may enhance the safety of present reactors, to the extend passive features are backfitted. This better understanding should also improve the safety of future reactors, which can incorporate more of these features. Passive safety systems and components may help to prevent accidents, core damage, or release radionuclides to the environment. The Technical Committee Meeting which was hosted by the USSR State Committee for Utilization of Nuclear Energy was attended by about 80 experts from 16 IAEA Member States and the NEA-OECD. A total of 21 papers were presented during the meeting. The objective of the meeting was to review and discuss passive safety systems and features of current and future water cooled reactor designs and to exchange information in this area of activity. A separate abstract was prepared for each of the 21 papers published in this proceedings. Refs, figs and tabs

  1. EPR design features to mitigate severe accident challenges

    International Nuclear Information System (INIS)

    Mazurkiewicz, S.M.; Fischer, M.; Bittermann, D.

    2005-01-01

    The EPR, an evolutionary pressurized water reactor (PWR), is a 4300-4500 MWth that incorporates proven technology within an optimized configuration to enhance safety. EPR was originally developed through a joint effort between Framatome ANP and Siemens by incorporating the best technological features from the French and German nuclear reactor fleets into a cost-competitive product. Commercial EPR units are currently being built in Finland at the Olkiluoto site, and planned for France at the Flamanville site. In recent months, Framatome ANP announced their intention to market the EPR units to China in response to a request for vendor bids as well as their intent to pursue design certification in the United States under 10CFR52. The EPR safety philosophy is based on a deterministic consideration of defense-in-depth complemented by probabilistic analyses. Not only is the EPR designed to prevent and mitigate design basis accidents (DBAs), it employs an extra level of safety associated with severe accident response. Therefore, as a design objective, features are included to ensure that radiological consequences are limited such that the need for stringent counter measures, such as evacuation and relocation of the nearby population, can be reasonably excluded. This paper discusses some of the innovative features of the EPR to address severe accident challenges. (author)

  2. The safety features of an integrated maritime reactor

    International Nuclear Information System (INIS)

    Miyakoshi, Junichi; Yamada, Nobuyuki; Kuwahara, Shin-ichi

    1975-01-01

    The EFDR-80, a typical integrated maritime reactor, which is being developed in West Germany is outlined. The safety features of the integrated maritime reactor are presented with the analysis of reactor accidents and hazards, and are compared with those of the separated maritime reactor. Furthermore, the safety criteria of maritime reactors in Japan and West Germany are compared, and some of the differences are presented from the viewpoint of reactor design and safety analysis. In this report the authors express an earnest desire that the definite and reasonable safety criteria of the integrated maritime reactor should be established and that the safety criteria of the nuclear ship should be standardized internationally. (auth.)

  3. The reactor safety study of experimental multi-purpose VHTR design

    International Nuclear Information System (INIS)

    Yasuno, T.; Mitake, S.; Ezaki, M.; Suzuki, K.

    1981-01-01

    Over the past years, the design works of the Experimental Very High Temperature Reactor (VHTR) plant have been conducted at Japan Atomic Energy Research Institute. The conceptual design has been completed and the more detailed design works and the safety analysis of the experimental VHTR plant are continued. The purposes of design studies are to show the feasibility of the experimental VHTR program, to specify the characteristics and functions of the plant components, to point out the R and D items necessary for the experimental VHTR plant construction, and to analyze the feature of the plant safety. In this paper the summary of system design and safety features of the experimental reactor are indicated. Main issues are the safety philosophy for the design basis accident, the accidents assumed and the engineered safety systems adopted in the design works

  4. Design features of APWR in Japan

    International Nuclear Information System (INIS)

    Yamaguchi, H.; Aeba, Y.; Weiss, E.H.

    1999-01-01

    Development of the Advanced Pressurized Water Reactor (APWR) was executed in the Improvement and Standardization Program which was organized by the Ministry of International Trade and Industry, Japanese utilities (Hokkaido, Kansai, Shikoku, Kyushu Electric Power Companies and the Japan Atomic Power Company) and manufacturers (Mitsubishi Heavy Industries and Westinghouse Electric). Improvements in terms of safety, reliability, operability, maintainability and economy have been incorporated based on comprehensive evaluations of both the advanced technologies available today, and the experience associated with construction and operation of current PWR plants. The main design improvement features applied in APWR include a core design that contributes to effective use of uranium resource, safety enhancement in the engineered safeguard system, and reliability enhancement in the reactor internal structures. This paper briefly describes the main features of the APWR design focusing on the following two items: the radial reflector, which enhances reliability of the reactor internal structures as well as neutron economy in the core region; and an advanced accumulator, which enhances Emergency Core Cooling System (ECCS) reliability and contributes to system simplification due to passive low pressure injection function. (author)

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

    International Nuclear Information System (INIS)

    Katanishi, Shoji; Kunitomi, Kazuhiko

    2003-01-01

    Japan Atomic Energy Research Institute has been developing design studies of the Gas Turbine High Temperature Reactor (GTHTR300). The original safety design philosophy has also been discussed and fixed for the GTHTR300. One of the unique feature of the safety philosophy of the GTHTR300 is that a depressurization accident is postulated as a design basis accident in order to show the high level of safety characteristics, though its probability of occurrence is much lower than the probability range of design basis accident. Another feature of safety design is to adopt a double confinement that is one of the original concepts for the GTHTR300. By using a double confinement, a feasibility of safety design without containment vessel was clarified even in case of a depressurization accident. This article describes the safety design philosophy and some results of preliminary evaluations which were conducted in order to clarify the feasibility of original safety design of the GTHTR300. (author)

  6. New source terms: what do they tell us about engineered safety feature performance

    International Nuclear Information System (INIS)

    Bernero, R.M.

    1985-01-01

    The accident behavior models which are the basis of engineered safety feature design are generally simple, non-mechanistic and concentrated on volatile radioiodine. Now data from source term studies show that models should be more mechanistic and look at other species than volatile iodine. A complete reevaluation of engineered safety features is needed

  7. Potential safety features and safety analysis aspects for high performance light water reactor (HPLWR)

    International Nuclear Information System (INIS)

    Aksan, N.; Schulenberg, T.; Squarer, D.

    2003-01-01

    Research Activities are ongoing worldwide to develop advanced nuclear power plants with high thermal efficiency for the purpose to improve their economical competitiveness. Within the 5th Framework Programme of the European Commission, a project has been launched with the main objective to assess the technical and economical feasibility of a high efficiency LWR operating at super critical pressure conditions. Several European research institutions, industrial partners and the University of Tokyo participated and worked in this common research project. Within the aims of the development of the HPLWR is to use both passive and active safety systems for performing safety related functions in the event of transients or accidents. Consequently substantial effort has been invested in order to define the safety features of the plant in a European environment, as well as to incorporate passive safety features into the design. Throughout this process, the European Utility Requirements (EUR) and requirements known from Generation IV initiative were considered as a guideline in general terms in order to include further advanced ideas. The HPLWR general features were compared to both requirements, indicating a potential to meet these. Since, the supercritical HPLWR represents a challenge for best-estimate safety codes like RELAP5, CATHARE and TRAB due to the fact that these codes were developed for two-phase or single-phase coolant at pressures far below critical point, work on the preliminary assessment of the appropriateness of these codes have been performed for selected relevant phenomena, and application of the codes to the selected transients on the basis of defined 'reference design'. An overview on their successful upgrade to supercritical pressures and application to some plant safety analysis are provided in the paper. Further elaborations in relation to future needs are also discussed. (author)

  8. SAFETY IN THE DESIGN OF SCIENCE LABORATORIES AND BUILDING CODES.

    Science.gov (United States)

    HOROWITZ, HAROLD

    THE DESIGN OF COLLEGE AND UNIVERSITY BUILDINGS USED FOR SCIENTIFIC RESEARCH AND EDUCATION IS DISCUSSED IN TERMS OF LABORATORY SAFETY AND BUILDING CODES AND REGULATIONS. MAJOR TOPIC AREAS ARE--(1) SAFETY RELATED DESIGN FEATURES OF SCIENCE LABORATORIES, (2) LABORATORY SAFETY AND BUILDING CODES, AND (3) EVIDENCE OF UNSAFE DESIGN. EXAMPLES EMPHASIZE…

  9. Innovative safety features of the modular HTGR

    International Nuclear Information System (INIS)

    Silady, F.A.; Simon, W.A.

    1992-01-01

    The Modular High Temperature Gas-Cooled Reactor (MHTGR) is an advanced reactor concept under development through a cooperative program involving the US Government, the nuclear industry, and the utilities. Near-term development is focused on electricity generation. The top-level safety requirement is that the plant's operation not disturb the normal day-to-day activities of the public. Quantitatively, this requires that the design meet the US Environmental Protection Agency's Protective Action Guides at the site boundary and hence preclude the need for sheltering or evacuation of the public. To meet these stringent safety requirements and at the same time provide a cost competitive design requires the innovative use of the basic high temperature gas-cooled reactor features of ceramic fuel, helium coolant, and a graphite moderator. The specific fuel composition and core size and configuration have been selected to the use the natural characteristics of these materials to develop significantly higher margins of safety. In this document the innovative safety features of the MHTGR are reviewed by examining the safety response to events challenging the functions relied on to retain radionuclides within the coated fuel particles. A broad range of challenges to core heat removal are examined, including a loss of helium pressure of a simultaneous loss of forced cooling of the core. The challenges to control of heat generation consider not only the failure to insert the reactivity control systems but also the withdrawal of control rods. Finally, challenges to control of chemical attack of the ceramic-coated fuel are considered, including catastrophic failure of the steam generator, which allows water ingress, or failure of the pressure vessels, which allows air ingress. The plant's response to these extreme challenges is not dependent on operator action, and the events considered encompass conceivable operator errors

  10. Safety design features for current UK advanced gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yellowlees, J. M.; Cobb, E. C. [Nuclear Power Co. (Risley) Ltd. (UK)

    1981-01-15

    The nuclear power stations planned for Heysham II and Torness will each have twin 660 MW(e) Advanced Gas-cooled Reactors (AGR) based on the design of those which have been operating at Hinkley Point 'B' and Hunterston 'B' since 1976. This paper has described the way in which the shutdown and cooling systems for the Heysham II and Torness AGRs have been selected in order to meet current UK safety requirements. Fault tree analyses have been used to identify the credible fault sequences, the probabilities of which have been calculated. By this means the relative importance of the various protective systems has been established and redundancy and reliability requirements identified. This systematic approach has led to a balanced design giving protection over the complete spectrum of fault sequences. Current safety requirements for thermal reactors in the UK and particular requirements in the design of the Heysham II and Torness reactors are discussed.

  11. Safety design features for current UK advanced gas-cooled reactors

    International Nuclear Information System (INIS)

    Yellowlees, J.M.; Cobb, E.C.

    1981-01-01

    The nuclear power stations planned for Heysham II and Torness will each have twin 660 MW(e) Advanced Gas-cooled Reactors (AGR) based on the design of those which have been operating at Hinkley Point 'B' and Hunterston 'B' since 1976. This paper has described the way in which the shutdown and cooling systems for the Heysham II and Torness AGRs have been selected in order to meet current UK safety requirements. Fault tree analyses have been used to identify the credible fault sequences, the probabilities of which have been calculated. By this means the relative importance of the various protective systems has been established and redundancy and reliability requirements identified. This systematic approach has led to a balanced design giving protection over the complete spectrum of fault sequences. Current safety requirements for thermal reactors in the UK and particular requirements in the design of the Heysham II and Torness reactors are discussed

  12. Integrated system of safety features for spent fuel interim storage

    International Nuclear Information System (INIS)

    Pantazi, Doina; Stanciu, Marcela; Mateescu, Silvia; Marin, Ion

    1999-01-01

    The design of the spent fuel interim storage facility (SFISF) must meet the applicable safety requirements in order to ensure radiological protection of the personnel, public and environment during all phases of the facility. To elaborate the safety documentation necessary for licensing, we were trying to chose the most appropriate approach related to safety features for SFISF, based on national and international regulations, standards and recommendations, as well as on the experience of other countries with similar facilities and finally, on our own experience in designing other nuclear objectives in Romania. The paper presents the issues that we consider important for the safety evaluation and are developed as a detailed diagram. The diagram contains in a logical succession the following issues: - fundamental principles of radioprotection; - fundamental safety principles of radioactive waste management; - safety objectives of SFISF; - safety criteria for SFISF; - safety requirements for SFISF; - siting criteria for SFISF; - siting requirements for SFISF. (authors)

  13. Design data and safety features of commercial nuclear power plants including cumulative index for Volumes I--VI

    International Nuclear Information System (INIS)

    Heddleson, F.A.

    1977-01-01

    Design data, safety features, and site characteristics are summarized for 12 nuclear power units in 6 power stations in the United States. Six pages of data are presented for each station, consisting of thermal-hydraulic and nuclear factors, containment features, emergency-core-cooling systems, site features, circulating water system data, and miscellaneous factors. In addition, an aerial perspective is presented for each plant. This volume covers plants with docket numbers 50-553 through 50-569 (Phipps Bend, Black Fox, Yellow Creek, and NEP) and two earlier plants not previously reported--Hope Creek (50-354, 50-355) and WPPSS 1 and 4 (50-460, 50-513). Indexes for this volume and the five earlier volumes are presented in three forms--by docket number, by plant name, and by participating utility

  14. Design characteristics of safety parameter display system for nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Yuangfang

    1992-02-01

    The design features of safety parameter display system (SPDS) developed by Tsinghua University is introduced. Some new features have been added into the system functions and they are: (1) hierarchical display structure; (2) human factor in the display format design; (3)automatic diagnosis of safety status of nuclear power plant; (4) extension of SPDS use scope; (5) flexible hardware structure. The new approaches in the design are: (1)adopting the international design standards; (2) selecting safety parameters strictly; (3) developing software under multitask operating system; (4) using a nuclear power plant simulator to verify the SPDS design

  15. Reactor design and safety approach for a tank-type fast reactor

    International Nuclear Information System (INIS)

    Davies, S.M.; Yamaki, Hideo; Goodman, L.

    1984-06-01

    A tank type plant has been designed that offers compactness, high reliability under seismic and thermal transients, and a safety design approach that provides a balance between public safety and plant availability. This report provides a description of the design philosophy and safety features of the reactor

  16. Passive and engineered safety features of the prototype fast reactor (PFR), Dounreay

    International Nuclear Information System (INIS)

    Gregory, C.V.

    1991-01-01

    Prototype fast reactor (PFR) combines passive and engineered safety features. Natural convection, a strong negative power coefficient, the decay heat removal system, and a fuel design able to operate beyond failure are all inherent and passive safety features of the PFR. The reliable shutdown system and the protection provided against SGU leaks are example of engineered protection. Experience at PFR demonstrates the worth and potential of a range of passive and engineered safeguards

  17. A Quantitative Feasibility Study on Potential Safety Improvement Effects of Advanced Safety Features in APR-1400 when Applied to OPR-1000

    Energy Technology Data Exchange (ETDEWEB)

    Ualikhan Zhiyenbayev [KAIST, Daejeon (Korea, Republic of); Chung, Dae Wook [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-10-15

    This study aims to test the feasibility of the applications using Probabilistic Safety Assessment (PSA). Particularly, three of those advanced safety features are selected as follows: 1. Providing an additional Emergency Diesel Generator (EDG); 2. Increasing the capacity of Class 1E batteries; 3. Placing a Refueling Water Storage Tank (RWST) inside containment, i.e., change from RWST to IRWST. The Advanced Power Reactor 1400 (APR-1400) adopts several advanced safety features compared to its predecessor, the Optimized Power Reactor 1000 (OPR-1000), which includes an additional Emergency Diesel Generator, increase in battery capacity, in-containment refueling water storage tank (IRWST), and so on. Considering the remarkable advantages of these safety features in safety improvement and the design similarities between APR-1400 and OPR-1000, it is feasible to apply key advanced safety features of APR-1400 to OPR-1000 to enhance the safety. The selected safety features are incorporated into OPR-1000 PSA model using the Advanced Information Management System (AIMS) for PSA and CDFs are re-evaluated for each application and combination of three applications. Based on current results, it is concluded that three of key advanced safety features of APR-1400 can be effectively applied to OPR-1000, resulting in considerable safety improvement. In aggregate, three advanced safety features, which are an additional EDG, increased battery capacity and IRWST, can reduce the CDF of OPR-1000 by more than 15% when applied altogether.

  18. A Quantitative Feasibility Study on Potential Safety Improvement Effects of Advanced Safety Features in APR-1400 when Applied to OPR-1000

    International Nuclear Information System (INIS)

    Ualikhan Zhiyenbayev; Chung, Dae Wook

    2015-01-01

    This study aims to test the feasibility of the applications using Probabilistic Safety Assessment (PSA). Particularly, three of those advanced safety features are selected as follows: 1. Providing an additional Emergency Diesel Generator (EDG); 2. Increasing the capacity of Class 1E batteries; 3. Placing a Refueling Water Storage Tank (RWST) inside containment, i.e., change from RWST to IRWST. The Advanced Power Reactor 1400 (APR-1400) adopts several advanced safety features compared to its predecessor, the Optimized Power Reactor 1000 (OPR-1000), which includes an additional Emergency Diesel Generator, increase in battery capacity, in-containment refueling water storage tank (IRWST), and so on. Considering the remarkable advantages of these safety features in safety improvement and the design similarities between APR-1400 and OPR-1000, it is feasible to apply key advanced safety features of APR-1400 to OPR-1000 to enhance the safety. The selected safety features are incorporated into OPR-1000 PSA model using the Advanced Information Management System (AIMS) for PSA and CDFs are re-evaluated for each application and combination of three applications. Based on current results, it is concluded that three of key advanced safety features of APR-1400 can be effectively applied to OPR-1000, resulting in considerable safety improvement. In aggregate, three advanced safety features, which are an additional EDG, increased battery capacity and IRWST, can reduce the CDF of OPR-1000 by more than 15% when applied altogether

  19. Technical specification optimization program - engineered safety features

    International Nuclear Information System (INIS)

    Andre, G.R.; Jansen, R.L.

    1986-01-01

    The Westinghouse Technical Specification Program (TOP) was designed to evaluate on a quantitative basis revisions to Nuclear Power Plant Technical Specifications. The revisions are directed at simplifying plant operation, and reducing unnecessary transients, shutdowns, and manpower requirements. In conjunction with the Westinghouse Owners Group, Westinghouse initiated a program to develop a methodology to justify Technical Specification revisions; particularly revisions related to testing and maintenance requirements on plant operation for instrumentation systems. The methodology was originally developed and applied to the reactor trip features of the reactor protection system (RPS). The current study further refined the methodology and applied it to the engineered safety features of the RPS

  20. The Alternative Design Features for Safety Enhancement in Shutdown Operation

    International Nuclear Information System (INIS)

    Oh, Hae Cheol; Kim, Myung Ki; Chung, Bag Soon; Seo, Mi Ro

    2009-01-01

    PSA can be used to confirm that the new plant design is complied with the applicable safety goals, and to select among the alternate design options. A shutdown PSA provides insight for outage planning schedule, outage management practices, and design modifications. Considering the results of both LPSD PSA studies and operating experiences for low power and shutdown, the improvements can be proposed to reduce the high risk contribution. The improvements/enhancements during shutdown operation may be divided into categories such as hardware, administrative management, and operational procedure. This paper presents on an example how the risk related to an accidental situation can be reduced, focusing the hardware design changes for the newly designed NPPs

  1. Safety features of subcritical fluid fueled systems

    International Nuclear Information System (INIS)

    Bell, C.R.

    1995-01-01

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

  2. Safety features of subcritical fluid fueled systems

    International Nuclear Information System (INIS)

    Bell, C.R.

    1994-01-01

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

  3. Safety features of subcritical fluid fueled systems

    Energy Technology Data Exchange (ETDEWEB)

    Bell, C.R. [Los Alamos National Laboratory, NM (United States)

    1995-10-01

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

  4. Safety of nuclear power plants: Design. Safety requirements

    International Nuclear Information System (INIS)

    2000-01-01

    other reactor types, including innovative developments in future systems, some of the requirements may not be applicable, or may need some judgment in their interpretation. Various Safety Guides will provide guidance in the interpretation and implementation of these requirements. This publication is intended for use by organizations designing, manufacturing, constructing and operating nuclear power plants as well as by regulatory bodies. This publication establishes design requirements for structures, systems and components important to safety that must be met for safe operation of a nuclear power plant, and for preventing or mitigating the consequences of events that could jeopardize safety. It also establishes requirements for a comprehensive safety assessment, which is carried out in order to identify the potential hazards that may arise from the operation of the plant, under the various plant states (operational states and accident conditions). The safety assessment process includes the complementary techniques of deterministic safety analysis and probabilistic safety analysis. These analyses necessitate consideration of postulated initiating events (PlEs), which include many factors that, singly or in combination, may affect safety and which may: originate in the operation of the nuclear power plant itself; be caused by human action; be directly related to the nuclear power plant and its environment. This publication also addresses events that are very unlikely to occur, such as severe accidents that may result in major radioactive releases, and for which it may be appropriate and practicable to provide preventive or mitigatory features in the design. This publication does not address: external natural or human induced events that are extremely unlikely (such as the impact of a meteorite or an artificial satellite); conventional industrial accidents that under no circumstances could affect the safety of the nuclear power plant; or non-radiological effects arising

  5. Safety of nuclear power plants: Design. Safety requirements

    International Nuclear Information System (INIS)

    2004-01-01

    other reactor types, including innovative developments in future systems, some of the requirements may not be applicable, or may need some judgment in their interpretation. Various Safety Guides will provide guidance in the interpretation and implementation of these requirements. This publication is intended for use by organizations designing, manufacturing, constructing and operating nuclear power plants as well as by regulatory bodies. This publication establishes design requirements for structures, systems and components important to safety that must be met for safe operation of a nuclear power plant, and for preventing or mitigating the consequences of events that could jeopardize safety. It also establishes requirements for a comprehensive safety assessment, which is carried out in order to identify the potential hazards that may arise from the operation of the plant, under the various plant states (operational states and accident conditions). The safety assessment process includes the complementary techniques of deterministic safety analysis and probabilistic safety analysis. These analyses necessitate consideration of postulated initiating events (PlEs), which include many factors that, singly or in combination, may affect safety and which may: originate in the operation of the nuclear power plant itself. Be caused by human action. Be directly related to the nuclear power plant and its environment. This publication also addresses events that are very unlikely to occur, such as severe accidents that may result in major radioactive releases, and for which it may be appropriate and practicable to provide preventive or mitigatory features in the design. This publication does not address: external natural or human induced events that are extremely unlikely (such as the impact of a meteorite or an artificial satellite). Conventional industrial accidents that under no circumstances could affect the safety of the nuclear power plant. Or non-radiological effects arising

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  8. Improved design features of KSNP+ BOP Fluid System

    International Nuclear Information System (INIS)

    Park, Heung Gyu; Yoon, Kyung Sup

    2002-01-01

    KOPEC (Korea Power Engineering Co.) in conjunction with the client KHNP (Korea Hydro and Nuclear Power Co.) has been developing the KSNP + (Improved Korean Standard Nuclear Power Plants) design concept since 1998. The main objective of the KSNP + is to enhance safety and economy of KSNP. The design concepts of the KSNP + will be implemented in Shin-Kori Units 1 and 2 Shin-Wolsung Units 1 and 2. This paper provides on an introduction to the improved design features of the KSNP + BOP fluid system consisting of 45 design improvement items. The design improvement concepts of the BOP fluid system have been developed as follows: optimization of system configuration and capacity, simplification of system, and adoption of advanced design features. Improved design features of the BOP fluid system allow additional benefits due to making a contribution to the optimization of plant arrangement and the reduction of operating costs during the plant life time. In conclusion, design improvement to the BOP fluid system have contributed to the KSNP + design concept being more reliable, safe and economically competitive

  9. The safety feature of hydraulic driving system of control rod for 200 MW nuclear heating reactor

    International Nuclear Information System (INIS)

    Chi Zongbo; Wu Yuanqiang

    1997-01-01

    The hydraulic driving system of control rod is used as control rod drive mechanism in 200 MW nuclear heating reactor. Design of this system is based on passive system, integrating drive and guide of control rod. The author analyzes the inherent safety and the design safety of this system, with mechanism of control rod not ejecting when the pressure of pressure vessel is lost, and calculating result of core not exposing when the amount of coolant is drained by broken pipe. The results indicate that this system has good safety feature, and assures reactor safety under any accident conditions, providing important technology support for 200 MW nuclear heating reactor with inherent safety feature

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

    International Nuclear Information System (INIS)

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

    1992-10-01

    Liquid metal cooled reactors (LMRs) have already been demonstrated to be robust machines. Many reactor designers now believe that it is possible to include in this technology sufficient passive safety that LMRs would be able to survive loss of flow, loss of heat sink, and transient overpower events, even if the plant protective system fails completely and do so without damage to the core. Early whole-core testing in Rapsodie, EBR-II. and FFTF indicate such designs may be possible. The operational safety testing program in EBR-II is demonstrating benign response of the reactor to a full range of controls failures. But additional testing is needed if transient core structural response under major accident conditions is to be properly understood. The proposed international Phase IIB passive safety tests in FFTF, being designed with a particular emphasis on providing, data to understand core bowing extremes, and further tests planned in EBR-11 with processed IFR fuel should provide a substantial and unique database for validating the computer codes being used to simulate postulated accident conditions

  11. The European space suit, a design for productivity and crew safety

    Science.gov (United States)

    Skoog, A. Ingemar; Berthier, S.; Ollivier, Y.

    In order to fulfil the two major mission objectives, i.e. support planned and unplanned external servicing of the COLUMBUS FFL and support the HERMES vehicle for safety critical operations and emergencies, the European Space Suit System baseline configuration incorporates a number of design features, which shall enhance the productivity and the crew safety of EVA astronauts. The work in EVA is today - and will be for several years - a manual work. Consequently, to improve productivity, the first challenge is to design a suit enclosure which minimizes movement restrictions and crew fatigue. It is covered by the "ergonomic" aspect of the suit design. Furthermore, it is also necessary to help the EVA crewmember in his work, by giving him the right information at the right time. Many solutions exist in this field of Man-Machine Interface, from a very simple system, based on cuff check lists, up to advanced systems, including Head-Up Displays. The design concept for improved productivity encompasses following features: • easy donning/doffing thru rear entry, • suit ergonomy optimisation, • display of operational information in alpha-numerical and graphical from, and • voice processing for operations and safety critical information. Concerning crew safety the major design features are: • a lower R-factor for emergency EVA operations thru incressed suit pressure, • zero prebreath conditions for normal operations, • visual and voice processing of all safety critical functions, and • an autonomous life support system to permit unrestricted operations around HERMES and the CFFL. The paper analyses crew safety and productivity criteria and describes how these features are being built into the design of the European Space Suit System.

  12. PHWR safety: design, siting and construction

    International Nuclear Information System (INIS)

    Sharma, V.K.

    2002-01-01

    In all activities associated with NPPs viz. siting, design, construction, commissioning and operation, safety is given overriding importance. The safety design principles of PHWRs are based on defence-in-depth approach, physical and functional separation between process and safety systems and also among various safety systems, redundancy to meet single failure criteria and postulation of a number of design basis events for which the plant must be designed. Apart from engineered safety systems, PHWRs have inherent characteristics which contribute to safety. In siting of a NPP, it is required to ensure that the given site does not pose undue radiological hazard to public and the environment both during normal operation as well as during and following an accident condition. For this purpose, all site related external events, both natural and man induced, are assessed for their effect on the plant and are considered as part of the design basis. Possible radiological impact of the NPP on environment and surrounding population is assessed and ensured to be within acceptable limits. During construction phase, it is essential that the NPP be built in accordance with design intent and with required quality of workmanship to ensure that the NPP will remain safe during all states of operation. This is achieved through careful execution and QA activities encompassing all aspects of component fabrication at manufacturer works, civil construction, site erection, assembly, and commissioning. Future trends in nuclear safety will continue to be based on existing principles which have proved to be sound. These will be further strengthened by features such as increasing use of passive means of performing safety functions and a more explicit treatment of severe accidents. (author)

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

    International Nuclear Information System (INIS)

    Katanishi, Shoji; Kunitomi, Kazuhiko

    2003-01-01

    Japan Atomic Energy Research Institute (JAERI) has been developing design studies of the Gas Turbine High Temperature Reactor (GTHTR300). The original safety design philosophy has also been discussed and fixed for the GTHTR300 based on the experience of the High Temperature Engineering Test Reactor (HTTR) of JAERI which is the first High Temperature Gas-cooled Reactor (HTGR) in Japan. One of the unique feature of the safety philosophy of the GTHTR300 is that a depressurization accident induced by a large pipe break is postulated as a design basis accident in order to show the high level of safety characteristics, though its probability of occurrence is lower than the probability range of design basis accident. Another feature of safety design is to adopt a double confinement that is one of the original concepts for the GTHTR300. By using a double confinement, a feasibility of safety design without containment vessel was clarified even in case of the depressurization accident. The safety design philosophies for passive cooling system, reactor shutdown system, and so on were determined. The methodology for the safety evaluation, such as safety criteria and selection of events to be evaluated by using estimation of probability of occurrence, were also discussed and determined. This article describes the safety design philosophy and some results of preliminary evaluations which were conducted in order to clarify the feasibility of original safety design of the GTHTR300. The present study is entrusted from Ministry of Education, Culture, Sports, Science and Technology of Japan. (author)

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

    International Nuclear Information System (INIS)

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

    2006-06-01

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

  15. On the status of the EFR Euro-Breeder and its passive safety features

    International Nuclear Information System (INIS)

    Marth, W.

    1992-01-01

    The Project of the EFR, the European Fast Reactor, is characterized by close European cooperation among power utilities, plant vendors, and research centers. In the present phase up until 1993 a consistent design of the nuclear part of the plant is being elaborated with the inclusion of a site-independent safety report. The most important design features, especially those in the field of passive safety, must be backed up by reliable R and D findings. These findings will enable the ad hoc Safety Club, a body of European safety experts, to pass its vote on the general licensability of the plant concept. (orig.) [de

  16. Safety considerations in the design of PFBR

    International Nuclear Information System (INIS)

    Vaidyanathan, G.; Om Pal Singh; Govindarajan, S.; Chellapandi, P.; Chetal, S.C.; Shankar Singh, R.; Bhoje, S.B.

    1996-01-01

    Prototype Fast Breeder Reactor (PFBR) is a 500 MWe reactor under design in India. The overall safety approach adopted is based on the defence-in-depth principle. Design features have been incorporated to minimize occurrence of unsafe conditions. A plant protection system comprising reliable core monitoring to detect the off-normal condition, a reliable shutdown system to ensure safe shutdown and a passive decay heat removal system are provided. Containment is provided to prevent any release of radioactivity to the environment in case of failure of the protective devices. This paper provides a brief outline of the safety considerations in the design of PFBR. (author). 5 refs, 1 tab

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

    International Nuclear Information System (INIS)

    2001-10-01

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

  18. Safety aspects of designs for future light water reactors (evolutionary reactors)

    International Nuclear Information System (INIS)

    1993-07-01

    The main purpose of this document is to describe the major innovations of proposed designs of future light water reactors, to describe specific safety characteristics and safety analysis methodologies, and to give a general overview of the most important safety aspects related to future reactors. The reactors considered in this report are limited to those intended for fixed station electrical power production, excluding most revolutionary concepts. More in depth discussion is devoted to those designs that are in a more advanced state of completion and have been more extensively described and analysed in the open literature. Other designs will be briefly described, as evidence of the large spectrum of new proposals. Some designs are similar; others implement unique features and require specific discussion (not all aspects of designs with unique features are fully discussed in this document). 131 refs, 22 figs

  19. DESIGN PACKAGE 1E SYSTEM SAFETY ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    M. Salem

    1995-06-23

    The purpose of this analysis is to systematically identify and evaluate hazards related to the Yucca Mountain Project Exploratory Studies Facility (ESF) Design Package 1E, Surface Facilities, (for a list of design items included in the package 1E system safety analysis see section 3). This process is an integral part of the systems engineering process; whereby safety is considered during planning, design, testing, and construction. A largely qualitative approach was used since a radiological System Safety Analysis is not required. The risk assessment in this analysis characterizes the accident scenarios associated with the Design Package 1E structures/systems/components(S/S/Cs) in terms of relative risk and includes recommendations for mitigating all identified risks. The priority for recommending and implementing mitigation control features is: (1) Incorporate measures to reduce risks and hazards into the structure/system/component design, (2) add safety devices and capabilities to the designs that reduce risk, (3) provide devices that detect and warn personnel of hazardous conditions, and (4) develop procedures and conduct training to increase worker awareness of potential hazards, on methods to reduce exposure to hazards, and on the actions required to avoid accidents or correct hazardous conditions.

  20. Safety assessment in plant layout design using indexing approach: implementing inherent safety perspective. Part 1 - guideword applicability and method description.

    Science.gov (United States)

    Tugnoli, Alessandro; Khan, Faisal; Amyotte, Paul; Cozzani, Valerio

    2008-12-15

    Layout planning plays a key role in the inherent safety performance of process plants since this design feature controls the possibility of accidental chain-events and the magnitude of possible consequences. A lack of suitable methods to promote the effective implementation of inherent safety in layout design calls for the development of new techniques and methods. In the present paper, a safety assessment approach suitable for layout design in the critical early phase is proposed. The concept of inherent safety is implemented within this safety assessment; the approach is based on an integrated assessment of inherent safety guideword applicability within the constraints typically present in layout design. Application of these guidewords is evaluated along with unit hazards and control devices to quantitatively map the safety performance of different layout options. Moreover, the economic aspects related to safety and inherent safety are evaluated by the method. Specific sub-indices are developed within the integrated safety assessment system to analyze and quantify the hazard related to domino effects. The proposed approach is quick in application, auditable and shares a common framework applicable in other phases of the design lifecycle (e.g. process design). The present work is divided in two parts: Part 1 (current paper) presents the application of inherent safety guidelines in layout design and the index method for safety assessment; Part 2 (accompanying paper) describes the domino hazard sub-index and demonstrates the proposed approach with a case study, thus evidencing the introduction of inherent safety features in layout design.

  1. Design provisions for safety

    International Nuclear Information System (INIS)

    Birkhofer, A.

    1983-01-01

    Design provisions for safety of nuclear power plants are based on a well balanced concept: the public is protected against a release of radioactive material by multiple barriers. These barriers are protected according to a 'defence-in-depth' principle. The reactor safety concept is primarily aimed at the prevention of accidents, especially fuel damage. Additionally, measures for consequence limitation are provided in order to prevent a severe release of radioactivity to the environment. However, it is difficult to judge the overall effectiveness of such devices. In a comprehensive safety analysis it has to be shown that the protection systems and safeguards work with sufficient reliability in the event of an accident. For the reliability assessment deterministic criteria (single failure, redundancy, fail-safe, demand for diversity) play an important role. Increasing efforts have been made to assess reliability quantitatively by means of probabilistic methods. It is now usual to perform reliability analyses of essential systems of nuclear power plants in the course of licensing procedures. As an additional level of emergency measures for a further reduction of hazards a reasonable amount of accident information has to be transferred. Operational experience may be considered as an important feedback to the design of plant safety features. Operator training has to include, besides skill in performing of operating procedures, the training of a flexible response to different accident situations. Experience has shown that the design provisions for safety could prevent dangerous release of the radioactive material to the environment after an accident has occurred. For future developments of reactor safety, extensive analyses of operating experience are of great importance. The main goal should be to enhance the reliability of measures for accident prevention, which prevent the core from meltdown or other damages

  2. HANARO cooling features: design and experience

    International Nuclear Information System (INIS)

    Park, Cheol; Chae, Hee-Taek; Han, Gee-Yang; Jun, Byung-Jin; Ahn, Guk-Hoon

    1999-01-01

    In order to achieve the safe core cooling during normal operation and upset conditions, HANARO adopted an upward forced convection cooling system with dual containment arrangements instead of the forced downward flow system popularly used in the majority of forced convection cooling research reactors. This kind of upward flow system was selected by comparing the relative merits of upward and downward flow systems from various points of view such as safety, performance, maintenance. However, several operational matters which were not regarded as serious at design come out during operation. In this paper are presented the design and operational experiences on the unique cooling features of HANARO. (author)

  3. The key design features of the Indian advanced heavy water reactor

    International Nuclear Information System (INIS)

    Sinha, R.K.; Kakodkar, A.; Anand, A.K.; Venkat Raj, V.; Balakrishnan, K.

    1999-01-01

    The 235 MWe Indian Advanced Heavy Water Reactor (AHWR) is a vertical, pressure tube type, boiling light water cooled reactor. The three key specific features of design of the AHWR, having a large impact on its viability, safety and economics, relate to its reactor physics, coolant channel, and passive safety features. The reactor physics design is tuned for maximising use of thorium based fuel, and achieving a slightly negative void coefficient of reactivity. The fulfilment of these requirements has been possible through use of PuO 2 -ThO 2 MOX, and ThO 2 -U 233 O 2 MOX in different pins of the same fuel cluster, and use of a heterogeneous moderator consisting of pyrolytic carbon and heavy water in 80%-20% volume ratio. The coolant channels of AHWR are designed for easy replaceability of pressure tubes, during normal maintenance shutdowns. The removal of pressure tube along with bottom end-fitting, using rolled joint detachment technology, can be done in AHWR coolant channels without disturbing the top end-fitting, tail pipe and feeder connections, and all other appendages of the coolant channel. The AHWR incorporates several passive safety features. These include core heat removal through natural circulation, direct injection of Emergency Core Coolant System (ECCS) water in fuel, passive systems for containment cooling and isolation, and availability of a large inventory of borated water in overhead Gravity Driven Water Pool (GDWP) to facilitate sustenance of core decay heat removal, ECCS injection, and containment cooling for three days without invoking any active systems or operator action. Incorporation of these features has been done together with considerable design simplifications, and elimination of several reactor grade equipment. A rigorous evaluation of feasibility of AHWR design concept has been completed. The economy enhancing aspects of its key design features are expected to compensate for relative complexity of the thorium fuel cycle activities

  4. Investigation of road network features and safety performance.

    Science.gov (United States)

    Wang, Xuesong; Wu, Xingwei; Abdel-Aty, Mohamed; Tremont, Paul J

    2013-07-01

    The analysis of road network designs can provide useful information to transportation planners as they seek to improve the safety of road networks. The objectives of this study were to compare and define the effective road network indices and to analyze the relationship between road network structure and traffic safety at the level of the Traffic Analysis Zone (TAZ). One problem in comparing different road networks is establishing criteria that can be used to scale networks in terms of their structures. Based on data from Orange and Hillsborough Counties in Florida, road network structural properties within TAZs were scaled using 3 indices: Closeness Centrality, Betweenness Centrality, and Meshedness Coefficient. The Meshedness Coefficient performed best in capturing the structural features of the road network. Bayesian Conditional Autoregressive (CAR) models were developed to assess the safety of various network configurations as measured by total crashes, crashes on state roads, and crashes on local roads. The models' results showed that crash frequencies on local roads were closely related to factors within the TAZs (e.g., zonal network structure, TAZ population), while crash frequencies on state roads were closely related to the road and traffic features of state roads. For the safety effects of different networks, the Grid type was associated with the highest frequency of crashes, followed by the Mixed type, the Loops & Lollipops type, and the Sparse type. This study shows that it is possible to develop a quantitative scale for structural properties of a road network, and to use that scale to calculate the relationships between network structural properties and safety. Copyright © 2013 Elsevier Ltd. All rights reserved.

  5. Safety Design Requirements for The Interior Architecture of Scientific Research Laboratories

    International Nuclear Information System (INIS)

    ElDib, A.A.

    2014-01-01

    The paper discusses one of the primary objectives of interior architecture design of research laboratories (specially those using radioactive materials) where it should provide a safe, accessible environment for laboratory personnel to conduct their work. A secondary objective is to allow for maximum flexibility for safe research. Therefore, health and safety hazards must be anticipated and carefully evaluated so that protective measures can be incorporated into the interior architectural design of these facilities wherever possible. The interior architecture requirements discussed in this paper illustrate some of the basic health and safety design features required for new and remodeled laboratories.The paper discusses one of the primary objectives of interior architecture design of research laboratories (specially those using radioactive materials) where it should provide a safe, accessible environment for laboratory personnel to conduct their work. A secondary objective is to allow for maximum flexibility for safe research. Therefore, health and safety hazards must be anticipated and carefully evaluated so that protective measures can be incorporated into the interior architectural design of these facilities wherever possible. The interior architecture requirements discussed in this paper illustrate some of the basic health and safety design features required for new and remodeled laboratories.

  6. Improved safety of the system 80+TM standard plants design through increased diversity and redundancy of safety systems

    International Nuclear Information System (INIS)

    Matzie, Regis A.; Carpentino, Frederick L.; Robertson, James E.

    1996-01-01

    Safely systems in the System 80+ TM Standard Plant are designed with more redundancy, diversity and simplicity than earlier nuclear power plant designs. These gains were accomplished by an evolutionary process that preserved the desirable and proven features in currently operating nuclear plants, while improving reliability and defense-in-depth. The System 80+ safety systems are the primary contributors to a core damage frequency that is more than 100 times lower than 1980's vintage U. S. designs, including the predecessor System 80 R standard nuclear steam supply system (NSSS) design. The System 80+ design includes significant improvements to the safety injection system, emergency feedwater system, shutdown cooling system, containment spray system, reactor coolant gas vent system, and to their vital support systems. These improvements enhance performance for traditional design basis events and significantly reduce the probability of a severe accident. The System 80+ design also incorporates safety systems to mitigate a severe accident. The added systems include the rapid depressurization system, the in-containment refueling water storage tank, the cavity flooding system. These systems fully address the U. S. Nuclear Regulatory Commission's (US NRC) severe accident policy. The System 80+ safety systems are integrated with the System 80+ Nuclear Island (NI) design. The NI general arrangement provides quadrant separation of the safety systems for protection from fire and flooding, and large equipment pull spaces and lay down areas for maintenance. This paper will describe the System 80+ safety systems advanced design features, the improved accident prevention and mitigation capabilities, and startup, operating and maintenance benefits

  7. Safety assessment in plant layout design using indexing approach: Implementing inherent safety perspective

    International Nuclear Information System (INIS)

    Tugnoli, Alessandro; Khan, Faisal; Amyotte, Paul; Cozzani, Valerio

    2008-01-01

    The design of layout plans requires adequate assessment tools for the quantification of safety performance. The general focus of the present work is to introduce an inherent safety perspective at different points of the layout design process. In particular, index approaches for safety assessment and decision-making in the early stages of layout design are developed and discussed in this two-part contribution. Part 1 (accompanying paper) of the current work presents an integrated index approach for safety assessment of early plant layout. In the present paper (Part 2), an index for evaluation of the hazard related to the potential of domino effects is developed. The index considers the actual consequences of possible escalation scenarios and scores or ranks the subsequent accident propagation potential. The effects of inherent and passive protection measures are also assessed. The result is a rapid quantification of domino hazard potential that can provide substantial support for choices in the early stages of layout design. Additionally, a case study concerning selection among various layout options is presented and analyzed. The case study demonstrates the use and applicability of the indices developed in both parts of the current work and highlights the value of introducing inherent safety features early in layout design

  8. AP1000 Containment Design and Safety Assessment

    International Nuclear Information System (INIS)

    Wright, Richard F.; Ofstun, Richard P.; Bachere, Sebastien

    2002-01-01

    The AP1000 is an up-rated version of the AP600 passive plant design that recently received final design certification from the US NRC. Like AP600, the AP1000 is a two-loop, pressurized water reactor featuring passive core cooling and passive containment safety systems. One key safety feature of the AP1000 is the passive containment cooling system which maintains containment integrity in the event of a design basis accident. This system utilizes a high strength, steel containment vessel inside a concrete shield building. In the event of a pipe break inside containment, a high pressure signal actuates valves which allow water to drain from a storage tank atop the shield building. Water is applied to the top of the containment shell, and evaporates, thereby removing heat. An air flow path is formed between the shield building and the containment to aid in the evaporation and is exhausted through a chimney at the top of the shield building. Extensive testing and analysis of this system was performed as part of the AP600 design certification process. The AP1000 containment has been designed to provide increased safety margin despite the increased reactor power. The containment volume was increased to accommodate the larger steam generators, and to provide increased margin for containment pressure response to design basis events. The containment design pressure was increased from AP600 by increasing the shell thickness and by utilizing high strength steel. The passive containment cooling system water capacity has been increased and the water application rate has been scaled to the higher decay heat level. The net result is higher margins to the containment design pressure limit than were calculated for AP600 for all design basis events. (authors)

  9. Distinctive safety aspects of the CANDU-PHW reactor design

    International Nuclear Information System (INIS)

    Kugler, G.

    1980-01-01

    Two lectures are presented in this report. They were prepared in response to a request from IAEA to provide information on the 'Special characteristics of the safety analysis of heavy water reactors' to delegates from member states attending the Interregional Training Course on Safety Analysis Review, held at Karlsruhe, November 19 to December 20, 1979. The CANDU-PHW reactor is used as a model for discussion. The first lecture describes the distinctive features of the CANDU reactor and how they impact on reactor safety. In the second lecture the Canadian safety philosophy, the safety design objective, and other selected topics on reactor safety analysis are discussed. The material in this report was selected with a view to assisting those not familiar with the CANDU heavy water reactor design in evaluating the distinctive safety aspects of these reactors. (auth)

  10. Major NSSS design features of the Korean next generation reactor

    International Nuclear Information System (INIS)

    Kim, Insk; Kim, Dong-Su

    1999-01-01

    In order to meet national needs for increasing electric power generation in the Republic of Korea in the 2000s, the Korean nuclear development group (KNDG) is developing a standardized evolutionary advanced light water reactor (ALWR), the Korean Next Generation Reactor (KNGR). It is an advanced version of the successful Korean Standard Nuclear Power Plant (KSNP) design, which meets utility needs for safety enhancement, performance improvement and ease of operation and maintenance. The KNGR design starts fro the proven design concept of the currently operating KSNPs with uprated power and advanced design features required by the utility. The KNGR design is currently in the final stage of the basic design, and the paper describes the major nuclear steam supply system (NSSS) design features of the KNGR together with introduction of the KNGR development program. (author)

  11. LFR safety features through intrinsic negative reactivity feedbacks

    International Nuclear Information System (INIS)

    Grasso, Giacomo

    2012-01-01

    The safety of Lead-cooled Fast Reactors can rely on intrinsic features such as: • the impossibility of Lead boiling, hence the unreliability of core (only) voiding; • the buoyancy of Control Rods in Lead, allowing their safe positioning also below the active region. For heightening the safety features of LFRs in safety analyses it could be required to approach the evaluation of the reactivity coefficients from a more physical point of view, including more elementary mechanisms, each one related to the proper driving temperature

  12. Passive safety design characteristics of the KALIMER-600 burner reactor

    International Nuclear Information System (INIS)

    Kwon, Young-Min; Jeong, Hae-Yong; Cho, Chung-Ho; Ha, Ki-Seok; Kim, Sang-Ji

    2009-01-01

    The Korea Atomic Energy Research Institute (KAERI) has recently studied several burner core designs for a transuranics (TRU) transmutation based on the breakeven core geometry of KALIMER-600. The KALIMER-600 is a net electrical rating of 600MWe, sodium-cooled, metallic-fueled, pool-type reactor. For the burner core concept selected for the present analysis, the smearing fractions of the fuel rods in three fuel zones are changed while maintaining the cladding outer diameter and cladding thickness. The resulting fuel slug smearing fractions of the inner, middle, and outer core zones are 36%, 40%, and 48%, respectively. The TRU conversion ratio is 0.57 and the TRU enrichment of the driver fuel is set to 30.0 w/o because of the current practical limitation of the U-TRU-10%Zr metal fuel database. The purpose of this paper is to evaluate the safety performance characteristics provided by the passive safety design features in the KALIMER-600 burner reactor by using a system-wide safety analysis code. The present scoping analysis focuses on an assessment of the enhanced safety design features that provide passive and self-regulating responses to transient conditions and an evaluation of the safety margin during unprotected overpower, unprotected loss of flow, and unprotected loss of heat sink events. The analysis results show that the KALIMER-600 burner reactor provides larger safety margins with respect to the sodium boiling, fuel rod integrity, and structural integrity. The overall inherent safety can be enhanced by accounting for the reactivity feedback mechanisms in the design process. (author)

  13. ARIES-AT safety design and analysis

    Energy Technology Data Exchange (ETDEWEB)

    Petti, D.A. [Idaho National Engineering and Environmental Laboratory, Fusion Safety Program, P.O. Box 1625, Idaho Falls, ID 83415 (United States)]. E-mail: David.Petti@inl.gov; Merrill, B.J. [Idaho National Engineering and Environmental Laboratory, Fusion Safety Program, P.O. Box 1625, Idaho Falls, ID 83415 (United States); Moore, R.L. [Idaho National Engineering and Environmental Laboratory, Fusion Safety Program, P.O. Box 1625, Idaho Falls, ID 83415 (United States); Longhurst, G.R. [Idaho National Engineering and Environmental Laboratory, Fusion Safety Program, P.O. Box 1625, Idaho Falls, ID 83415 (United States); El-Guebaly, L. [Fusion Technology Institute, 1500 Engineering Drive, University of Wisconsin-Madison, Madison, WI 53706 (United States); Mogahed, E. [Fusion Technology Institute, 1500 Engineering Drive, University of Wisconsin-Madison, Madison, WI 53706 (United States); Henderson, D. [Fusion Technology Institute, 1500 Engineering Drive, University of Wisconsin-Madison, Madison, WI 53706 (United States); Wilson, P. [Fusion Technology Institute, 1500 Engineering Drive, University of Wisconsin-Madison, Madison, WI 53706 (United States); Abdou, A. [Fusion Technology Institute, 1500 Engineering Drive, University of Wisconsin-Madison, Madison, WI 53706 (United States)

    2006-01-15

    ARIES-AT is a 1000 MWe conceptual fusion power plant design with a very low projected cost of electricity. The design contains many innovative features to improve both the physics and engineering performance of the system. From the safety and environmental perspective, there is greater depth to the overall analysis than in past ARIES studies. For ARIES-AT, the overall spectrum of off-normal events to be examined has been broadened. They include conventional loss of coolant and loss of flow events, an ex-vessel loss of coolant, and in-vessel off-normal events that mobilize in-vessel inventories (e.g., tritium and tokamak dust) and bypass primary confinement such as a loss of vacuum and an in-vessel loss of coolant with bypass. This broader examination of accidents improves the robustness of the design from the safety perspective and gives additional confidence that the facility can meet the no-evacuation requirement under average weather conditions. We also provide a systematic assessment of the design to address key safety functions such as confinement, decay heat removal, and chemical energy control. In the area of waste management, both the volume of the component and its hazard are used to classify the waste. In comparison to previous ARIES designs, the overall waste volume is less because of the compact design.

  14. Safety parameter display system (SPDS) for Russian-designed NPPs

    International Nuclear Information System (INIS)

    Anikanov, S.S.; Catullo, W.J.; Pelusi, J.L.

    1997-01-01

    As part of the programs aimed at improving the safety of Russian-designed reactors, the US DoE has sponsored a project of providing a safety parameter display system (SPDS) for nuclear power plants with such reactors. The present paper is focused mostly on the system architecture design features of SPDS systems for WWER-1000 and RBMK-1000 reactors. The function and the operating modes of the SPDS are outlined, and a description of the display system is given. The system architecture and system design of both an integrated and a stand-alone IandC system is explained. (A.K.)

  15. Advanced gas cooled reactors - Designing for safety

    International Nuclear Information System (INIS)

    Keen, Barry A.

    1990-01-01

    The Advanced Gas-Cooled Reactor Power Stations recently completed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station output of 2x660 MW(e) gross although powers in excess of this have been achieved and it is currently intended to uprate this as far as possible. The design of both stations has been based on the successful operating AGRs at Hinkley Point and Hunterston which have now been in-service for almost 15 years, although minor changes were made to meet new safety requirements and to make improvements suggested by operating experience. The construction of these new AGRs has been to programme and within budget. Full commercial load for the first reactor at Torness was achieved in August 1988 with the other three reactors following over the subsequent 15 months. This paper summarises the safety principles and guidelines for the design of the reactors and discusses how some of the main features of the safety case meet these safety requirements. The paper also summarises the design problems which arose during the construction period and explains how these problems were solved with the minimum delay to programme

  16. Advanced gas cooled reactors - Designing for safety

    Energy Technology Data Exchange (ETDEWEB)

    Keen, Barry A [Engineering Development Unit, NNC Limited, Booths Hall, Knutsford, Cheshire (United Kingdom)

    1990-07-01

    The Advanced Gas-Cooled Reactor Power Stations recently completed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station output of 2x660 MW(e) gross although powers in excess of this have been achieved and it is currently intended to uprate this as far as possible. The design of both stations has been based on the successful operating AGRs at Hinkley Point and Hunterston which have now been in-service for almost 15 years, although minor changes were made to meet new safety requirements and to make improvements suggested by operating experience. The construction of these new AGRs has been to programme and within budget. Full commercial load for the first reactor at Torness was achieved in August 1988 with the other three reactors following over the subsequent 15 months. This paper summarises the safety principles and guidelines for the design of the reactors and discusses how some of the main features of the safety case meet these safety requirements. The paper also summarises the design problems which arose during the construction period and explains how these problems were solved with the minimum delay to programme.

  17. Passive safety features for next generation CANDU power plants

    International Nuclear Information System (INIS)

    Natalizio, A.; Hart, R.S.; Lipsett, J.J.; Soedijono, P.; Dick, J.E.

    1989-01-01

    CANDU offers an evolutionary approach to simpler and safer reactors. The CANDU 3, an advanced CANDU, currently in the detailed design stage, offers significant improvements in the areas of safety, design simplicity, constructibility, operability, maintainability, schedule and cost. These are being accomplished by retaining all of the well known CANDU benefits, and by relying on the use of proven components and technologies. A major safety benefit of CANDU is the moderator system which is separate from the coolant. The presence of a cold moderator reduces the consequences arising from a LOCA or loss of heat sink event. In existing CANDU plants even the severe accident - LOCA with failure of the emergency core cooling system - is a design basis event. Further advances toward a simpler and more passively safe reactor will be made using the same evolutionary approach. Building on the strength of the moderator system to mitigate against severe accidents, a passive moderator cooling system, depending only on the law of gravity to perform its function, will be the next step of development. AECL is currently investigating a number of other features that could be incorporated in future evolutionary CANDU designs to enhance protection against accidents, and to limit off-site consequences to an acceptable level, for even the worst event. The additional features being investigated include passive decay heat removal from the heat transport system, a simpler emergency core cooling system and a containment pressure suppression/venting capability for beyond design basis events. Central to these passive decay heat removal schemes is the availability of a short-term heat sink to provide a decay heat removal capability of at least three days, without any station services. Preliminary results from these investigations confirm the feasibility of these schemes. (author)

  18. Safety design approach for JSFR toward the realization of GEN IV SFR

    International Nuclear Information System (INIS)

    Kubo, S.; Yamano, H.; Chikazawa, Y.; Shimakawa, Y.

    2013-01-01

    Conclusion: Safety Design Approach for JSFR: • Based on the safety design criteria for Generation-IV SFR • DECs, Situations practically eliminated and related design measures are identified and selected with due consideration of the safety features of SFR and the lessons learned from the TEPCO’s Fukushima Dai-ichi nuclear power plants accident Safety Design Concept of JSFR: • For failure to shutdown: Passive shutdown capability, Mitigation of core damage (Prevention of severe mechanical energy release, In-Vessel Retention) • For failure to remove heat: Prevention of significant core damage (Natural circulation DHR, Alternative cooling measures) • Containment: Prevention of sever dynamic loads by design measures (IVR, double boundary concept, inertization)

  19. Structural health monitoring feature design by genetic programming

    International Nuclear Information System (INIS)

    Harvey, Dustin Y; Todd, Michael D

    2014-01-01

    Structural health monitoring (SHM) systems provide real-time damage and performance information for civil, aerospace, and other high-capital or life-safety critical structures. Conventional data processing involves pre-processing and extraction of low-dimensional features from in situ time series measurements. The features are then input to a statistical pattern recognition algorithm to perform the relevant classification or regression task necessary to facilitate decisions by the SHM system. Traditional design of signal processing and feature extraction algorithms can be an expensive and time-consuming process requiring extensive system knowledge and domain expertise. Genetic programming, a heuristic program search method from evolutionary computation, was recently adapted by the authors to perform automated, data-driven design of signal processing and feature extraction algorithms for statistical pattern recognition applications. The proposed method, called Autofead, is particularly suitable to handle the challenges inherent in algorithm design for SHM problems where the manifestation of damage in structural response measurements is often unclear or unknown. Autofead mines a training database of response measurements to discover information-rich features specific to the problem at hand. This study provides experimental validation on three SHM applications including ultrasonic damage detection, bearing damage classification for rotating machinery, and vibration-based structural health monitoring. Performance comparisons with common feature choices for each problem area are provided demonstrating the versatility of Autofead to produce significant algorithm improvements on a wide range of problems. (paper)

  20. Performance and safety design of the advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Berglund, R.C.; Magee, P.M.; Boardman, C.E.; Gyorey, G.L.

    1991-01-01

    The Advanced Liquid Metal Reactor (ALMR) program led by General Electric is developing, under U.S. Department of Energy sponsorship, a conceptual design for an advanced sodium-cooled liquid metal reactor plant. This design is intended to improve the already excellent level of plant safety achieved by the nuclear power industry while at the same time providing significant reductions in plant construction and operating costs. In this paper, the plant design and performance are reviewed, with emphasis on the ALMR's unique passive design safety features and its capability to utilize as fuel the actinides in LWR spent fuel

  1. Cost-benefit evaluation of containment related engineered safety features of Indian pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Bajaj, S.S.; Bhawal, R.N.; Rustagi, R.S.

    1984-01-01

    The typical containment system for a commercial nuclear reactor uses several engineered safety features to achieve its objective of limiting the release of radioactive fission products to the environment in the event of postulated accident conditions. The design of containment systems and associated features for Indian Pressurized Heavy Water Reactors (PHWRs) has undergone progressive improvement in successive projects. In particular, the current design adopted for the Narora Atomic Power Project (NAPP) has seen several notable improvements. The paper reports on a cost-benefit study in respect of three containment related engineered safety features and subsystems of NAPP, viz. (i) secondary containment envelope, (ii) primary containment filtration and pump-back system, and (iii) secondary containment filtration, recirculation and purge system. The effect of each of these systems in reducing the environmental releases of radioactivity following a design basis accident is presented. The corresponding reduction in population exposure and the associated monetary value of this reduction in exposure are also given. The costs of the features and subsystem under consideration are then compared with the monetary value of the exposures saved, as well as other non-quantified benefits, to arrive at conclusions regarding the usefulness of each subsystem. This study clearly establishes for the secondary containment envelope the benefit in terms of reduction in public exposure giving a quantitative justification for the costs involved. In the case of the other two subsystems, which involve relatively low costs, while all benefits have not been quantified, their desirability is justified on qualitative considerations. It is concluded that the engineered safety features adopted in the current containment system design of Indian PHWRs contribute to reducing radiation exposures during accident conditions in accordance with the ALARA ('as low as reasonably achievable') principle

  2. Novel modular natural circulation BWR design and safety evaluation

    International Nuclear Information System (INIS)

    Ishii, Mamoru; Shi, Shanbin; Yang, Won Sik; Wu, Zeyun; Rassame, Somboon; Liu, Yang

    2015-01-01

    Highlights: • Introduction of BWR-type natural circulation small modular reactor preliminary design (NMR-50). • Design of long fuel cycle length for the NMR-50. • Design of double passive safety systems for the NMR-50. • RELAP5 analyses of design basis accidents for the NMR-50. - Abstract: The Purdue NMR (Novel Modular Reactor) represents a BWR-type small modular reactor with a significantly reduced reactor pressure vessel (RPV) height. Specifically, it has one third the height of a conventional BWR RPV with an electrical output of 50 MWe. The preliminary design of the NMR-50 including reactor, fuel cycle, and safety systems is described and discussed. The improved neutronics design of the NMR-50 extends the fuel cycle length up to 10 years. The NMR-50 is designed with double passive engineering safety system, which is intended to withstand a prolonged station black out with loss of ultimate heat sink accident such as experienced at Fukushima. In order to evaluate the safety features of the NMR-50, two representative design basis accidents, i.e. main steam line break (MSLB) and bottom drain line break (BDLB), are simulated by using the best-estimate thermal–hydraulic code RELAP5. The RPV water inventory, containment pressure, and the performance of engineering safety systems are investigated for about 33 h after the initiation of the accidents

  3. Safety of Nuclear Power Plants: Design. Specific Safety Requirements

    International Nuclear Information System (INIS)

    2012-01-01

    On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  4. License Application Design Selection Feature Report: Waste Package Self Shielding Design Feature 13

    International Nuclear Information System (INIS)

    Tang, J.S.

    2000-01-01

    In the Viability Assessment (VA) reference design, handling of waste packages (WPs) in the emplacement drifts is performed remotely, and human access to the drifts is precluded when WPs are present. This report will investigate the feasibility of using a self-shielded WP design to reduce the radiation levels in the emplacement drifts to a point that, when coupled with ventilation, will create an acceptable environment for human access. This provides the benefit of allowing human entry to emplacement drifts to perform maintenance on ground support and instrumentation, and carry out performance confirmation activities. More direct human control of WP handling and emplacement operations would also be possible. However, these potential benefits must be weighed against the cost of implementation, and potential impacts on pre- and post-closure performance of the repository and WPs. The first section of this report will provide background information on previous investigations of the self-shielded WP design feature, summarize the objective and scope of this document, and provide quality assurance and software information. A shielding performance and cost study that includes several candidate shield materials will then be performed in the subsequent section to allow selection of two self-shielded WP design options for further evaluation. Finally, the remaining sections will evaluate the impacts of the two WP self-shielding options on the repository design, operations, safety, cost, and long-term performance of the WPs with respect to the VA reference design

  5. Design-related inherent safety characteristics in large LMFBR power plants

    International Nuclear Information System (INIS)

    Tzanos, C.P.; Barthold, W.P.; Bowers, C.H.; Ferguson, D.R.; Prohammer, F.G.; van Erp, J.B.

    1976-01-01

    Design-related safety-enhancing features such as (1) extended pump coastdown, (2) increased negative reactivity feedbacks, (3) reduced sodium void reactivity, and (4) self-actuated shutdown systems are evaluated. Primary emphasis is placed on preventing or limiting core damage. Attention is also given to features aimed at mitigation of the energetics potential of hypothetical core-disruptive accidents

  6. Design, fabrication and erection of steel structures important to safety of nuclear facilities

    International Nuclear Information System (INIS)

    2001-10-01

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

  7. Inherent Safety Feature of Hybrid Low Power Research Reactor during Reactivity Induced Accident

    Energy Technology Data Exchange (ETDEWEB)

    Kim, DongHyun; Yum, Soo Been; Hong, Sung Teak; Lim, In-Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Hybrid low power research reactor(H-LPRR) is the new design concept of low power research reactor for critical facility as well as education and training. In the case of typical low power research reactor, the purposes of utilization are the experiments for education of nuclear engineering students, Neutron Activation Analysis(NAA) and radio-isotope production for research purpose. H-LPRR is a light-water cooled and moderated research reactor that uses rod-type LEU UO{sub 2} fuels same as those for commercial power plants. The maximum core thermal power is 70kW and, the core is placed in the bottom of open pool. There are 1 control rod and 2 shutdown rods in the core. It is designed to cool the core by natural convection, retain negative feedback coefficient for entire fuel periods and operate for 20 years without refueling. Inherent safety in H-LPRR is achieved by passive design features such as negative temperature feedback coefficient and core cooling by natural convection during normal and emergency conditions. The purpose of this study is to find out that the inherent safety characteristics of H-LPRR is able to control the power and protect the reactor from the RIA(Reactivity induced accident). RIA analysis was performed to investigate the inherent safety feature of H-LPRR. As a result, it was found that the reactor controls its power without fuel damage in the event and that the reactor remains safe states inherently. Therefore, it is believed that high degree of safety inheres in H-LPRR.

  8. Enhanced CANDU 6 design assist probabilistic safety assessment results and insights

    International Nuclear Information System (INIS)

    Torabi, T.; Bettig, R.; Iliescu, P.; Robinson, J.; Santamaura, P.; Skorupska, B.; Tyagi, A.K.; Vencel, I.

    2013-01-01

    The Enhanced CANDU 6(EC6) is a 700 MWe reactor, which has evolved from the well-established CANDU line of reactors, which are heavy-water moderated, and heavy-water cooled horizontal pressure tube reactors, using natural uranium fuel. The EC6 design retains the generic CANDU design features, while incorporating innovations and state-of-the-art technologies to ensure competitiveness with other design with respect to operation, performance and economics. A design assist probabilistic safety assessment (PSA) was conducted during the design change phase of the project. The purpose of the assessment was to assess internal events during at-power operation and identify the design improvements and additional features needed to comply with the latest regulatory requirements in Canada and compete with other reactor designs, internationally. The PSA results show that the EC6 plant response to the postulated initiating events is well balanced, and the design meets its safety objectives. This paper summarizes the results and insights gained during the development of the PSA models for at-power internal events. (author)

  9. Innovative safety features of the modular HTGR

    International Nuclear Information System (INIS)

    Silady, F.A.; Simon, W.A.

    1992-04-01

    In this document the innovative safety features of the MHTGR are reviewed by examining the safety response to events challenging the functions relied on to retain radionuclides within the coated fuel particles. A broad range of challenges to core heat removal are examined, including a loss of helium pressure and a simultaneous loss of forced cool of the core

  10. Risk assessment of safety data link and network communication in digital safety feature control system of nuclear power plant

    International Nuclear Information System (INIS)

    Lee, Sang Hun; Son, Kwang Seop; Jung, Wondea; Kang, Hyun Gook

    2017-01-01

    Highlights: • Safety data communication risk assessment framework and quantitative scheme were proposed. • Fault-tree model of ESFAS unavailability due to safety data communication failure was developed. • Safety data link and network risk were assessed based on various ESF-CCS design specifications. • The effect of fault-tolerant algorithm reliability of safety data network on ESFAS unavailability was assessed. - Abstract: As one of the safety-critical systems in nuclear power plants (NPPs), the Engineered Safety Feature-Component Control System (ESF-CCS) employs safety data link and network communication for the transmission of safety component actuation signals from the group controllers to loop controllers to effectively accommodate various safety-critical field controllers. Since data communication failure risk in the ESF-CCS has yet to be fully quantified, the ESF-CCS employing data communication systems have not been applied in NPPs. This study therefore developed a fault tree model to assess the data link and data network failure-induced unavailability of a system function used to generate an automated control signal for accident mitigation equipment. The current aim is to provide risk information regarding data communication failure in a digital safety feature control system in consideration of interconnection between controllers and the fault-tolerant algorithm implemented in the target system. Based on the developed fault tree model, case studies were performed to quantitatively assess the unavailability of ESF-CCS signal generation due to data link and network failure and its risk effect on safety signal generation failure. This study is expected to provide insight into the risk assessment of safety-critical data communication in a digitalized NPP instrumentation and control system.

  11. Innovative features and fuel design approach in the iris reactor

    International Nuclear Information System (INIS)

    Petrovic, B.; Carelli, M.; Greenspan, E.; Matsumoto, H.; Padovani, E.; Ganda, F.

    2002-01-01

    The International Reactor Innovative and Secure (IRIS) is being developed by an international consortium of industry, laboratory, university and utility establishments, led by Westinghouse. The IRIS design addresses key requirements associated with advanced reactors, including improved safety, enhanced proliferation resistance, competitive electricity production cost, and improved waste management. IRIS is a modular, small/medium size (335 MWe) PWR with an integral vessel configuration. The objective has been to base its design on proven LWR technology, so that no new technology development is needed and near-term deployment is possible, yet at the same time to introduce innovative features making it attractive when compared to present PWRs. These opposing requirements resulted in an evolutionary approach to fuel and core design, balancing new features against the need to avoid extensive testing and demonstration programmes. (author)

  12. Design features to achieve defence-in-depth in small and medium sized reactors

    International Nuclear Information System (INIS)

    Kuznetsov, Vladimir

    2009-01-01

    Broader incorporation of inherent and passive safety design features has become a 'trademark' of many advanced reactor concepts, including several evolutionary designs and nearly all innovative small and medium sized design concepts. Ensuring adequate defence-in-depth is important for reactors of smaller output because many of them are being designed to allow more proximity to the user, specifically, when non-electrical energy products are targeted. Based on the activities recently performed by the International Atomic Energy Agency, the paper provides a summary description of the design features used to achieve defence in depth in the eleven representative concepts of small and medium sized reactors. (author)

  13. Integrating Safeguards and Security with Safety into Design

    International Nuclear Information System (INIS)

    Bean, Robert S.; Hockert, John W.; Hebditch, David J.

    2009-01-01

    There is a need to minimize security risks, proliferation hazards, and safety risks in the design of new nuclear facilities in a global environment of nuclear power expansion, while improving the synergy of major design features and raising operational efficiency. In 2008, the U.S. Department of Energy (DOE), National Nuclear Security Administration (NNSA) launched the Next Generation Safeguards Initiative (NGSI) covering many safeguards areas. One of these, launched by NNSA with support of the DOE Office of Nuclear Energy, was a multi-laboratory project, led by the Idaho National Laboratory (INL), to develop safeguards by design. The proposed Safeguards-by-Design (SBD) process has been developed as a structured approach to ensure the timely, efficient, and cost effective integration of international safeguards and other nonproliferation barriers with national material control and accountability, physical security, and safety objectives into the overall design process for the nuclear facility lifecycle. A graded, iterative process was developed to integrate these areas throughout the project phases. It identified activities, deliverables, interfaces, and hold points covering both domestic regulatory requirements and international safeguards using the DOE regulatory environment as exemplar to provide a framework and guidance for project management and integration of safety with security during design. Further work, reported in this paper, created a generalized SBD process which could also be employed within the licensed nuclear industry and internationally for design of new facilities. Several tools for integrating safeguards, safety, and security into design are discussed here. SBD appears complementary to the EFCOG TROSSI process for security and safety integration created in 2006, which focuses on standardized upgrades to enable existing DOE facilities to meet a more severe design basis threat. A collaborative approach is suggested.

  14. The Safety Feature Test of QNX RTOS

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jang Yeol; Lee, Young Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    Benchmarking is a point of reference by which something can be measured. The QNX is a kind of Real Time Operating System(RTOS) developed by QSSL(QNX Software Systems Ltd.) in Canada. The ELMSYS is the brand name of commercially available PC to be applied such as Cabinet Operator Module(COM) of Digital Plant Protection System(DPPS) and COM of Digital Engineered Safety Features Actuation System(DESFAS-AC). The ELMSYS PC Hardware will be qualified by KTL(Korea Testing Lab.) in order to use as a Cabinet Operator Module(COM). QNX RTOS is dedicating by KAERI now. This paper describes the outline and some safety features among benchmarking test for QNX RTOS under the ELMSYS PC platform

  15. The Safety Feature Test of QNX RTOS

    International Nuclear Information System (INIS)

    Kim, Jang Yeol; Lee, Young Jun

    2010-01-01

    Benchmarking is a point of reference by which something can be measured. The QNX is a kind of Real Time Operating System(RTOS) developed by QSSL(QNX Software Systems Ltd.) in Canada. The ELMSYS is the brand name of commercially available PC to be applied such as Cabinet Operator Module(COM) of Digital Plant Protection System(DPPS) and COM of Digital Engineered Safety Features Actuation System(DESFAS-AC). The ELMSYS PC Hardware will be qualified by KTL(Korea Testing Lab.) in order to use as a Cabinet Operator Module(COM). QNX RTOS is dedicating by KAERI now. This paper describes the outline and some safety features among benchmarking test for QNX RTOS under the ELMSYS PC platform

  16. Hualong One's nuclear reactor core design and relative safety issues research

    Energy Technology Data Exchange (ETDEWEB)

    Yu, H., E-mail: yuhong_xing@126.com [Nuclear Power Inst. of China, Design and Research Sub-Inst., Chengdu, Sichuan (China)

    2015-07-01

    'Full text:' Hualong One, a third generation 1000MWe-class pressurized water reactor, is developed by China National Nuclear Cooperation (CNNC), based on the self-reliant technologies and experiences from China 40 years designing, construction, operation and maintenance of NPPs. In China, it has been approved to construct at Fuqing 5&6 and Fangchenggang 3&4. The Hualong One adopts advanced design features to dramatically enhance plant safety, economic efficiency and convenience of operation and maintenance. It consists of three loops with nominal thermal power output 3060 MWt and a 60-year design life. Its reactor core has 177 fuel assemblies, 18 month refueling interval (after initial cycle), and more than 15% thermal margin. It adopts low leakage loading pattern which can achieve better economy of the neutron, higher reactivity and lower radiation damage of pressure vessel. For the safety design, incorporating the feedback of Fukushima accident, the Hualong One has a combination of active and passive safety systems, a single station layout, double containment structure, and comprehensive implementation of defence-in-depth design principles. The new design features has been successfully evaluated to ensure that they enhance the performance and safety of Hualong One. Several experimental activates have been conducted, such as cavity injection and cooling system testing, passive containment heat removal system testing, and passive residual heat removal system of secondary side testing. The future improvements of Hualong reactor will focus on better economic core design and more reliable safety system. (author)

  17. Accommodation of unprotected accidents by inherent safety design features in metallic and oxide-fueled LMFBRs

    International Nuclear Information System (INIS)

    Cahalan, J.E.; Sevy, R.H.; Su, S.F.

    1985-01-01

    This paper presents the results of a study of the effectivness of intrinsic design features to mitigate the consequences of unprotected accidents in metallic and oxide-fueled LMFBRs. The accidents analyzed belong to the class generally considered to lead to core disruption; unprotected loss-of-flow (LOF) and transient over-power (TOP). Results of the study demonstrate the potential for design features to meliorate accident consequences, and in some cases to render them benign. Emphasis is placed on the relative performance of metallic and oxide-fueled core designs

  18. Research and development on reduced-moderation light water reactor with passive safety features (Contract research)

    International Nuclear Information System (INIS)

    Iwamura, Takamichi; Okubo, Tsutomu; Akie, Hiroshi; Kugo, Teruhiko; Yonomoto, Taisuke; Kureta, Masatoshi; Ishikawa, Nobuyuki; Nagaya, Yasunobu; Araya, Fumimasa; Okajima, Shigeaki; Okumura, Keisuke; Suzuki, Motoe; Mineo, Hideaki; Nakatsuka, Toru

    2004-06-01

    The present report contains the achievement of 'Research and Development on Reduced-moderation Light Water Reactor with Passive Safety Features', which was performed by Japan Atomic Energy Research Institute (JAERI), Hitachi Ltd., Japan Atomic Power Company and Tokyo Institute of Technology in FY2000-2002 as the innovative and viable nuclear energy technology (IVNET) development project operated by the Institute of Applied Energy (IAE). In the present project, the reduced-moderation water reactor (RMWR) has been developed to ensure sustainable energy supply and to solve the recent problems of nuclear power and nuclear fuel cycle, such as economical competitiveness, effective use of plutonium and reduction of spent fuel storage. The RMWR can attain the favorable characteristics such as high burnup, long operation cycle, multiple recycling of plutonium (Pu) and effective utilization of uranium resources based on accumulated LWR technologies. Our development target is 'Reduced-moderation Light Water Reactor with Passive Safety Features' with innovative technologies to achieve above mentioned requirement. Electric power is selected as 300 MWe considering anticipated size required for future deployment. The reactor core consists of MOX fuel assemblies with tight lattice arrangement to increase the conversion ratio. Design targets of the core specification are conversion ratio more than unity, negative void reactivity feedback coefficient to assure safety, discharged burnup more than 60 GWd/t and operation cycle more than 2 years. As for the reactor system, a small size natural circulation BWR with passive safety systems is adopted to increase safety and reduce construction cost. The results obtained are as follows: As regards core design study, core design was performed to meet the goal. Sequence of startup operation was constructed for the RMWR. As the plant design, plant system was designed to achieve enhanced economy using passive safety system effectively. In

  19. A Technique of Software Safety Analysis in the Design Phase for PLC Based Safety-Critical Systems

    International Nuclear Information System (INIS)

    Koo, Seo-Ryong; Kim, Chang-Hwoi

    2017-01-01

    The purpose of safety analysis, which is a method of identifying portions of a system that have the potential for unacceptable hazards, is firstly to encourage design changes that will reduce or eliminate hazards and, secondly, to conduct special analyses and tests that can provide increased confidence in especially vulnerable portions of the system. For the design and implementation phase of the PLC based systems, we proposed a technique for software design specification and analysis, and this technique enables us to generate software design specifications (SDSs) in nuclear fields. For the safety analysis in the design phase, we used architecture design blocks of NuFDS to represent the architecture of the software. On the basis of the architecture design specification, we can directly generate the fault tree and then use the fault tree for qualitative analysis. Therefore, we proposed a technique of fault tree synthesis, along with a universal fault tree template for the architecture modules of nuclear software. Through our proposed fault tree synthesis in this work, users can use the architecture specification of the NuFDS approach to intuitively compose fault trees that help analyze the safety design features of software.

  20. SWR 1000: an advanced boiling water reactor with passive safety features

    International Nuclear Information System (INIS)

    Brettschuh, W.

    1999-01-01

    The SWR 1000, an advanced BWR, is being developed by Siemens under contract from Germany's electric utilities and with the support of European partners. The project is currently in the basic design phase to be concluded in mid-1999 with the release of a site-independent safety report and costing analysis. The development goals for the project encompass competitive costs, use of passive safety systems to further reduce probabilities of occurrence of severe accidents, assured control of accidents so no emergency response actions for evacuation of the local population are needed, simplification of plant systems based on operator experience, and planning and design based on German codes, standards and specifications put forward by the Franco-German Reactor Safety Commission for future nuclear power plants equipped with PWRs, as well as IAEA specifications and the European Utility Requirements. These goals led to a plant concept with a low power density core, with large water inventories stored above the core inside the reactor pressure vessel, in the pressure suppression pool, and in other locations. All accident situations arising from power operation can be controlled by passive safety features without rise in core temperature and with a grace period of more than three days. In addition, postulated core melt is controlled by passive equipment. All new passive systems have been successfully tested for function and performance using large-scale components in experimental testing facilities at PSI in Switzerland and at the Juelich Research Centre in Germany. In addition to improvements of the safety systems, the plant's operating systems have been simplified based on operating experience. The design's safety concept, simplified operating systems and 48 months construction time yield favourable plant construction costs. The level of concept maturity required to begin offering the SWR 1000 on the power generation market is anticipated to be reached, as planned in the year

  1. Passive safety features of low sodium void worth metal fueled cores in a bottom supported reactor vessel

    International Nuclear Information System (INIS)

    Chang, Y.I.; Marchaterre, J.F.; Wade, D.C.; Wigeland, R.A.; Kumaoka, Yoshio; Suzuki, Masao; Endo, Hiroshi; Nakagawa, Hiroshi

    1991-01-01

    A study has been performed on the passive safety features of low-sodium-void-worth metallic-fueled reactors with emphasis on using a bottom-supported reactor vessel design. The reactor core designs included self-sufficient types as well as actinide burners. The analyses covered the reactor response to the unprotected, i.e. unscrammed, transient overpower accident and the loss-of-flow accident. Results are given demonstrating the safety margins that were attained. 4 refs., 4 figs., 2 tabs

  2. SWR 1000: the main design features of the advanced boiling water reactor with passive safety systems

    International Nuclear Information System (INIS)

    Carsten, Pasler

    2007-01-01

    The SWR-1000 (1000 MW) is a boiling water reactor whose economic efficiency in comparison with large-capacity designs is achieved by deploying very simple passive safety equipment, simplified systems for plant operation, and a very simple plant configuration in which systems engineering is optimized and dependence on electrical and instrumentation and control systems is reduced. In addition, systems and components that require protection against natural and external man-made hazards are accommodated in such a way that as few buildings as possible have to be designed to withstand the loads from such events. The fuel assemblies have been enlarged from a 10*10 rod array to a 12*12 array. This reduces the total number of fuel assemblies in the core and thus also the number of control rods and control rod drives, as well as in-core neutron flux monitors. The design owes its competitiveness to the fact that investment costs, maintenance costs and fuel cycle costs are all lower. In addition, refueling outages are shorter, thanks to the reduced scope of outage activities. The larger fuel assemblies have been extensively and successfully tested, as have all of the other new components and systems incorporated into the plant design. As in existing plants, the forced coolant circulation method is deployed, ensuring problem-free startup, and enabling plant operators to adjust power rapidly in the high power range (70%-100%) without moving the control rods, as well as allowing spectral-shift and stretch-out operation. The plant safety concept is based on a combination of passive safety systems and a reduced number of active safety systems. All postulated accidents can be controlled using passive systems alone. Control of a postulated core melt accident is assured with considerable safety margins thanks to passive flooding of the containment for in-vessel melt retention. The SWR-1000 is compliant with international nuclear codes and standards, and is also designed to withstand

  3. Cost reduction and safety design features of new nuclear power plants in India. Annex 13

    International Nuclear Information System (INIS)

    Sharma, V.K.

    2002-01-01

    Indian Nuclear Power Programme is designed to exploit limited reserves of uranium and extensive resource of thorium. Pressurised heavy water reactors are found most suitable and form the main stay of the first stage of the programme. Thorium utilisation is achieved in the second and third stages. Today India has total installed capacity of 2720 MWe of PHWRs which are operating with high plant load factors of over 80%. Rich experience of construction and operation of over 150 reactor years is being utilised in effecting cost reduction and safety improvements. Standardisation and reduction in gestation period by preproject activities, advance procurement and work packages of engineer, procure, construct and commission are some of the techniques being adopted for cost reduction in the new projects. But the cost of safety is rising. Design basis event of double ended guillotine rupture of primary pressure boundary needs a relook based on current knowledge of material behaviour. This event appears improbable. Similarly some of the safety related systems like closed loop cooling water operating at low temperature and pressure, and low usage factors may be designed as per standard codes without invoking special nuclear requirements. The paper will address these issues and highlight the possible areas for cost reduction both in operating and safety systems. Modern construction and project management techniques are being employed. Gestation period of 5 years and cost of less than US $1400 per KWe are the present targets. In Indian environment nuclear power is found to be competitive with thermal power plants at distances of about 800 Kms from the coal mines. (author)

  4. Advances in global development and deployment of small modular reactors and incorporating lessons learned from the Fukushima Daiichi accident into the designs of engineered safety features of advanced reactors

    International Nuclear Information System (INIS)

    Hadid Subki, M.; )

    2014-01-01

    The IAEA has been facilitating the Member States in incorporating the lessons-learned from the Fukushima Dai-ichi Accident into the designs of engineered safety features of advanced reactors, including small modular reactors. An extended assessment is required to address challenges for advancing reactor safety in the new evolving generation of SMR plants to preserve the historic lessons in safety, through: assuring the diversity in emergency core cooling systems following loss of onsite AC power; ensuring diversity in reactor depressurization following a transient or accident; confirming independence in reactor trip and safety systems for sensors, power supplies and actuation systems, and finally diversity in maintaining containment integrity following a severe accident

  5. Applicability of object-oriented design methods and C++ to safety-critical systems

    International Nuclear Information System (INIS)

    Cuthill, B.B.

    1994-01-01

    This paper reports on a study identifying risks and benefits of using a software development methodology containing object-oriented design (OOD) techniques and using C++ as a programming language relative to selected features of safety-critical systems development. These features are modularity, functional diversity, removing ambiguous code, traceability, and real-time performance

  6. SGHWR safety design and evaluation

    International Nuclear Information System (INIS)

    Smith, D.R.; Merrett, D.J.; Ward, D.A.

    1977-01-01

    The paper discusses the characteristic features of the S.G.H.W.R. and identifies the single channel concept as of considerable importance. The unique feature of the design is the provision of individual spray cooling E.C.C.S. to each channel. This spray cooling occupies a prominent position in the main line safety arguments. The reliance on this form of spray cooling leads to provision of a comprehensive E.C.C.S. system of high reliability. Duplicate systems with diverse power and water sources cover the complete pressure range to give very high confidence that spray cooling is available in all major L.O.C.A.s. On the other hand hydraulic analysis of the blowdown phase demonstrates that significant convective flow is available as an alternative/supplementary cooling regime for most faults. The reactor shutdown mechanisms have also been duplicated and will be designed to high reliabilities to give surety of reactor trip in all credible faults. The comparative performance of the two systems is considered. Extent of diversity and redundancy in trip parameters is also discussed. A feature of channel concept is that the pipe sizes can be made relatively small thus restricting rates of blowdown, and the paper discusses effects of this upon long term cooling and flooding arguments. The quantities of pipework in the primary circuit introduce considerations of integrity and the paper goes on to list the measures introduced to improve segregation and protection of individual sections of the plant so that the extent of possible L.O.C.A.s is minimised. The achievement of high standards of reliability by use of in-service inspection is covered, with particular reference to the steam drums. The impact of these inspection requirements upon the very low man-rem exposures required by U.K. utilities is also included. Finally, it is noted that the provision of containment in common with other L.W.R. practice also provides a valuable engineered safety feature. The principles of

  7. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Chinese Ed.)

    International Nuclear Information System (INIS)

    2012-01-01

    On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  8. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (French Ed.)

    International Nuclear Information System (INIS)

    2012-01-01

    On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  9. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Arabic Ed.)

    International Nuclear Information System (INIS)

    2012-01-01

    On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  10. Development of design and safety analysis supporting system for casks

    International Nuclear Information System (INIS)

    Ohsono, Katsunari; Higashino, Akira; Endoh, Shuji

    1993-01-01

    Mitsubishi heavy Industries has developed a design and safety analysis supporting system 'CADDIE' (Cask Computer Aided Design, Drawing and Integrated Evaluation System), with the following objectives: (1) Enhancement of efficiency of the design and safety analysis (2) Further advancement of design quality (3) Response to the diversification of design requirements. The features of this system are as follows: (1) The analysis model data common to analyses is established, and it is prepared automatically from the model made by CAD. (2) The input data for the analysis code is available by simple operation of conversation type from the analysis model data. (3) The analysis results are drawn out in diagrams by output generator, so as to facilitate easy observation. (4) The data of material properties, fuel assembly data, etc. required for the analyses are made available as a data base. (J.P.N.)

  11. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Spanish Edition)

    International Nuclear Information System (INIS)

    2012-01-01

    This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  12. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Russian Edition)

    International Nuclear Information System (INIS)

    2012-01-01

    This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  13. Small nuclear reactor safety design requirements for autonomous operation

    International Nuclear Information System (INIS)

    Kozier, K.S.; Kupca, S.

    1997-01-01

    Small nuclear power reactors offer compelling safety advantages in terms of the limited consequences that can arise from major accident events and the enhanced ability to use reliable, passive means to eliminate their occurrence by design. Accordingly, for some small reactor designs featuring a high degree of safety autonomy, it may be-possible to delineate a ''safety envelope'' for a given set of reactor circumstances within which safe reactor operation can be guaranteed without outside intervention for time periods of practical significance (i.e., days or weeks). The capability to operate a small reactor without the need for highly skilled technical staff permanently present, but with continuous remote monitoring, would aid the economic case for small reactors, simplify their use in remote regions and enhance safety by limiting the potential for accidents initiated by inappropriate operator action. This paper considers some of the technical design options and issues associated with the use of small power reactors in an autonomous mode for limited periods. The focus is on systems that are suitable for a variety of applications, producing steam for electricity generation, district heating, water desalination and/or marine propulsion. Near-term prospects at low power levels favour the use of pressurized, light-water-cooled reactor designs, among which those having an integral core arrangement appear to offer cost and passive-safety advantages. Small integral pressurized water reactors have been studied in many countries, including the test operation of prototype systems. (author)

  14. HTGR safety philosophy

    Energy Technology Data Exchange (ETDEWEB)

    Joksimovic, V.; Fisher, C. R. [General Atomic Co., San Diego, CA (USA)

    1981-01-15

    The accident at the Three Mile Island has focused public attention on reactor safety. Many public figures advocate a safer method of generating nuclear electricity for the second nuclear era in the U.S. The paper discusses the safety philosophy of a concept deemed suitable for this second nuclear era. The HTGR, in the course of its evolution, included safety as a significant determinant in design philosophy. This is particularly evident in the design features which provide inherent safety. Inherent features cause releases from a wide spectrum of accident conditions to be low. Engineered features supplement inherent features. The significance of HTGR safety features is quantified and order-of-magnitude type of comparisons are made with alternative ways of generating electricity.

  15. HTGR safety philosophy

    International Nuclear Information System (INIS)

    Joksimovic, V.; Fisher, C.R.

    1981-01-01

    The accident at the Three Mile Island has focused public attention on reactor safety. Many public figures advocate a safer method of generating nuclear electricity for the second nuclear era in the U.S. The paper discusses the safety philosophy of a concept deemed suitable for this second nuclear era. The HTGR, in the course of its evolution, included safety as a significant determinant in design philosophy. This is particularly evident in the design features which provide inherent safety. Inherent features cause releases from a wide spectrum of accident conditions to be low. Engineered features supplement inherent features. The significance of HTGR safety features is quantified and order-of-magnitude type of comparisons are made with alternative ways of generating electricity. (author)

  16. HTGR safety philosophy

    International Nuclear Information System (INIS)

    Joskimovic, V.; Fisher, C.R.

    1980-08-01

    The accident at the Three Mile Island has focused public attention on reactor safety. Many public figures advocate a safer method of generating nuclear electricity for the second nuclear era in the US. The paper discusses the safety philosophy of a concept deemed suitable for this second nuclear era. The HTGR, in the course of its evolution, included safety as a significant determinant in design philosophy. This is particularly evident in the design features which provide inherent safety. Inherent features cause releases from a wide spectrum of accident conditions to be low. Engineered features supplement inherent features. The significance of HTGR safety features is quantified and order-of-magnitude type of comparisons are made with alternative ways of generating electricity

  17. Safety Features of Material and Personnel Movement Devices. Module SH-25. Safety and Health.

    Science.gov (United States)

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

    This student module on safety features of material and personnel movement devices is one of 50 modules concerned with job safety and health. This module covers safe conditions and operating practices for conveyors, elevators, escalators, moving walks, manlifts, forklifts, and motorized hand trucks. Following the introduction, 10 objectives (each…

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

    International Nuclear Information System (INIS)

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

    1976-10-01

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

  19. ITER final design report, cost review and safety analysis (FDR) and relevant documents

    International Nuclear Information System (INIS)

    1999-01-01

    This volume contains the fourth major milestone report and documents associated with its acceptance, review and approval. This ITER Final Design Report, Cost Review and Safety Analysis was presented to the ITER Council at its 13th meeting in February 1998 and was approved at its extraordinary meeting on 25 June 1998. The contents include an outline of the ITER objectives, the ITER parameters and design overview as well as operating scenarios and plasma performance. Furthermore, design features, safety and environmental characteristics and schedule and cost estimates are given

  20. Evaluation of special safety features of the SNR-300 in view of the Chernobyl accident

    International Nuclear Information System (INIS)

    Vossebrecker, H.

    1987-03-01

    A comparison of those characteristics, which decisively influenced the accident in the RMBK-1000 reactor, with the safety features of SNR-300 has been performed. The conclusions of this comparison are presented in the present report. The SNR-300 is characterized by a stable reactivity behaviour and good controllability, whereas RBMK-1000 has an instable behaviour and complex spatial dependencies in the core. Among other points, design deficiencies in the protection and emergency shutdown systems were responsible for the Chernobyl accident. The protection and scram systems of the SNR-300 are unquestionably superior to those of the RBMK-1000 with regard to redundancy, diversity, degree of automation, separation of operational and safety-relevant tasks, protection against inadmissible interventions, effectiveness and safety reserves. Therefore, excursion accidents can be classified as hypothetical for SNR-300. Due to elementary physical properties, possible energy releases during hypothetical excursions are substantially lower for SNR-300 and would be controlled by the design of the primary system and containment systems. No damage limiting measures are provided in the RBMK-100 for excursion accidents. Finally, exothermal processes augmented the consequences of the accident in the RBMK-1000 and the long-lasting graphite fire intensified the release of radioactivity. In the SNR-300, however, inertisation of the containment, the steel plate lining and the floor troughs ensure that activity enclosure inside the containment after leakage or hypothetical excursion accident is not endangered by exothermal reactions. Further safety aspects are presented in the report, which can be linked with the accident in Chernobyl. In summary, it is obvious that the disadvantageous physical and technical features of the RBMK-1000 do either not exist in the SNR-300 or are covered by the safety design

  1. USNRC regulatory guidance for engineered safety feature air cleaning systems

    International Nuclear Information System (INIS)

    Bellamy, R.R.

    1991-01-01

    The need for clear, technically appropriate, and easily implementable guidance for the design, testing, and maintenance of nuclear air cleaning systems has long been recognized. Numerous industry consensus standards have been issued and revised over the last 30 years. Guidance has also been published by the US Nuclear Regulatory Commission in the form of regulations, regulatory guides, standard review plans, NUREG documents, and information notices. This paper will summarize the latest revisions to these documents and emphasize Regulatory Guide 1.52, Design, Testing, and Maintenance Criteria for Post-Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants, which was last revised in 1978. The USNRC has undertaken a project to revise this regulatory guide, and the status of that revision is highlighted

  2. Key features of MIR.1200 (AES-2006) design and current stage of Leningrad NPP-2 construction

    International Nuclear Information System (INIS)

    Ivkov, Igor

    2010-01-01

    MIR.1200/AES-2006 is an abbreviated name of the evolving NPP design developed on the basis of the VVER-1000 Russian design with gross operation life of 480 reactor-years. This design is being implemented in four Units of Leningrad NPP-2 (LNPP-2. The AES-91/99 was used as reference during development of the AES-2006 design for LNPP-2; this design was implemented in two Units of Tianwan NPP (China). The main technical features of the MIR.1200/AES-2006 design include a double containment, four trains of active safety systems (4x100%, 4x50%), and special engineering measures for BDBA management (core catcher, H2 PARs, PHRS) based mainly on passive principles. The containment is described in detail, the main features in comparison with the reference NPP are outlined, the design layout principles are highlighted, the safety system structure and parameters are described. Attention is paid to the BDBA management system, hydrogen removal system, core catcher, and PHRS-SG and C-PHRS. (P.A.)

  3. System Design and the Safety Basis

    International Nuclear Information System (INIS)

    Ellingson, Darrel

    2008-01-01

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

  4. Advancements in the design of safety-related systems and components of the MARS nuclear plant

    International Nuclear Information System (INIS)

    Caira, M.; Caruso, G.; Naviglio, A.; Sorabella, L.; Farello, C.E.

    1992-01-01

    In the paper, the advancements in the design of safety-related systems and components of the MARS nuclear plant, equipped with a 600 MW th PWR, are described. These advancements are due to the special safety features of this plant, which relies completely on inherent and passive safety. In particular, the new steps of the design of the innovative, completely passive, and with an unlimited autonomy Emergency core Cooling System are described, together with the characteristics of the last version of the steam generator, developed in a new design involving disconnecting components, for a fast erection and an easy maintenance. (author)

  5. DART - for design basis justification and safety related information management

    International Nuclear Information System (INIS)

    Billington, A.; Blondiaux, P.; Boucau, J.; Cantineau, B.; Doumont, C.; Mared, A.

    2000-01-01

    DART is the acronym for Design Analysis Re-engineering Tool. It embodies a systematic and integrated approach to NPP safety re-assessment and configuration management, that makes use of Reverse Failure Mode and Effect Analysis in conjunction with a state-of-the-art relational database and a standardized data format, to permit long-term management of plant safety related information. The plant design is reviewed in a step-by-step logical fashion by constructing fault trees that identify the link between undesired consequences and their causes. Each failure cause identified in a fault tree is addressed by defining functional requirements, which are in turn addressed by documenting the specific manner in which the plant complies with the requirement. The database can be used to generate up-to-date plant safety related documents, including: SAR, Systems Descriptions, Technical Specifications and plant procedures. The approach is open-minded by nature and therefore is not regulatory driven, however the plant licensing basis will also be reviewed and documented within the same database such that a Regulatory Conformance Program may be integrated with the other safety documentation. This methodology can thus reconstitute the plant design bases in a comprehensive and systematic way, while allowing to uncover weaknesses in design. The original feature of the DART methodology is that it links all the safety related documents together, facilitating the evaluation of the safety impact resulting from any plant modification. Due to its capability to retrieve the basic justifications of the plant design, it is also a useful tool for training the young generation of plant personnel. The DART methodology has been developed for application to units 2, 3 and 4 at Vattenfall's Ringhals site in Sweden. It may be applied to any nuclear power plant or industrial facility where public safety is a concern. (author)

  6. DART - for design basis justification and safety related information management

    International Nuclear Information System (INIS)

    Billington, A.; Blondiaux, B.; Boucau, J.; Cantineau, B.; Mared, A.

    2001-01-01

    DART is the acronym for Design Analysis Re-Engineering Tool. It embodies a systematic and integrated approach to NPP safety re-assessment and configuration management, that makes use of Reverse Failure Mode and Effect Analysis in conjunction with a state-of-the-art relational database and a standardized data format, to permit long-term management of plant safety related information. The plant design is reviewed in a step-by-step logical fashion by constructing fault trees that identify the link between undesired consequences and their causes. Each failure cause identified in a fault tree is addressed by defining functional requirements, which are in turn addressed by documenting the specific manner in which the plant complies with the requirement. The database can then be used to generate up-to-date plant safety related documents, including: SAR, Systems Descriptions, Technical Specifications and plant procedures. The approach is open-minded by nature and therefore is not regulatory driven, however the plant licensing basis will also be reviewed and documented within the same database such that a Regulatory Conformance Program may be integrated with the other safety documentation. This methodology can thus reconstitute the plant design bases in a comprehensive and systematic way, while allowing to uncover weaknesses in design. The original feature of the DART methodology is that it links all the safety related documents together, facilitating the evaluation of the safety impact resulting from any plant modification. Due to its capability to retrieve the basic justifications of the plant design, it is also a useful tool for training the young generation of plant personnel. The DART methodology has been developed for application to units 2, 3 and 4 at Vattenfall's Ringhals site in Sweden. It may be applied to any nuclear power plant or industrial facility where public safety is a concern. (author)

  7. Critical safety issues in the design of fusion machines

    International Nuclear Information System (INIS)

    Kramer, W.

    1991-01-01

    In the course of developing fusion machines both general safety considerations and safety assessments for the various components and systems of actual machines increase in number and become more and more coherent. This is particularly true for the NET/ITER projects where safety analysis plays an increasing role for the design of the machine. Since in a D/T tokamak the radiological hazards will be dominant basic radiological safety objectives are discussed. Critical safety issues as identified in particular by the NET/ITER community are reviewed. Subsequently, issues of major concern are considered both for normal operation and for conceivable accidents. The following accidents are considered to be crucial: Loss of cooling in plasma facing components, loss of vacuum, tritium system failure, and magnet system failure. To mitigate accident consequences a confinement concept based on passive features and multiple barriers including detritiation and filtering has to be applied. The reactor building as final barrier needs special attention to cope with both internal and external hazards. (orig.)

  8. Review of research in feature based design

    NARCIS (Netherlands)

    Salomons, O.W.; van Houten, Frederikus J.A.M.; Kals, H.J.J.

    1993-01-01

    Research in feature-based design is reviewed. Feature-based design is regarded as a key factor towards CAD/CAPP integration from a process planning point of view. From a design point of view, feature-based design offers possibilities for supporting the design process better than current CAD systems

  9. Reliability study: digital engineered safety feature actuation system of Korean Standard Nuclear Power Plant

    International Nuclear Information System (INIS)

    Sudarno; Kang, H. G.; Jang, S. C.; Eom, H. S.; Ha, J. J.

    2003-04-01

    The usage of digital Instrumentation and Control (I and C) in a nuclear power plant becomes more extensive, including safety related systems. The PSA application of these new designs are very important in order to evaluate their reliability. In particular, Korean Standard Nuclear Power Plants (KSNPPs), typically Ulchin 5 and 6 (UCN 5 and 6) reactor units, adopted the digital safety-critical systems such as Digital Plant Protection System (DPPS) and Digital Engineered Safety Feature Actuation System (DESFAS). In this research, we developed fault tree models for assessing the unavailability of the DESFAS functions. We also performed an analysis of the quantification results. The unavailability results of different DESFAS functions showed that their values are comprised from 5.461E-5 to 3.14E-4. The system unavailability of DESFAS AFAS-1 is estimated as 5.461E-5, which is about 27% less than that of analog system if we consider the difference of human failure probability estimation between both analyses. The results of this study could be utilized in risk-effect analysis of KSNPP. We expect that the safety analysis result will contribute to design feedback

  10. Integrated Safety in Design

    DEFF Research Database (Denmark)

    Schultz, Casper Siebken; Jørgensen, Kirsten

    2014-01-01

    An on-going research project investigates the inclusion of health and safety considerations in the design phase as a means to achieve a higher level of health and safety in the construction industry. Moreover, the approach is coupled to the overall quality efforts. Two architectural firms and two...... consulting engineering firms are project participants. The hypothesis is that health and safety problems in execution can be prevented through better planning in the early stages of the construction processes and that accidents are prevented by providing safety. In the first stage of the research project...... a theoretical framework is developed from a combination of existing literature on health and safety and a mapping of existing practices based on interviews in all four companies. The interviews revealed that the basic knowledge on OHS among architects and engineers is limited. Also currently designers typically...

  11. Impact of Passive Safety on FHR Instrumentation Systems Design and Classification

    International Nuclear Information System (INIS)

    Holcomb, David Eugene

    2015-01-01

    operation of systems and components important to safety as required in GDC 20. This paper provides an overview of the design process employed to develop a pre-conceptual FHR instrumentation architecture intended to lower plant capital and operational costs by minimizing reliance on expensive, safety related, safety-significant instrumentation through the use of inherent passive features of FHRs.

  12. Recent Development of the Two-Stroke Engine. II - Design Features. 2; Design Features

    Science.gov (United States)

    Zeman, J.

    1945-01-01

    Completing the first paper dealing with charging methods and arrangements, the present paper discusses the design forms of two-stroke engines. Features which largely influence piston running are: (a) The shape and surface condition of the sliding parts. (b) The cylinder and piston materials. (c) Heat conditions in the piston, and lubrication. There is little essential difference between four-stroke and two-stroke engines with ordinary pistons. In large engines, for example, are always found separately cast or welded frames in which the stresses are taken up by tie rods. Twin piston and timing piston engines often differ from this design. Examples can be found in many engines of German or foreign make. Their methods of operation will be dealt with in the third part of the present paper, which also includes the bibliography. The development of two-stroke engine design is, of course, mainly concerned with such features as are inherently difficult to master; that is, the piston barrel and the design of the gudgeon pin bearing. Designers of four-stroke engines now-a-days experience approximately the same difficulties, since heat stresses have increased to the point of influencing conditions in the piston barrel. Features which notably affect this are: (a) The material. (b) Prevailing heat conditions.

  13. Safety aspects of LMR [liquid metal-cooled reactor] core design

    International Nuclear Information System (INIS)

    Cahalan, J.E.

    1986-01-01

    Features contributing to increased safety margins in liquid metal-cooled reactor (LMR) design are identified. The technical basis is presented for the performance of a pool-type reactor system with an advanced metallic alloy fuel in unprotected accidents. Results are presented from analyses of anticipated transients without scram, including loss-of-flow (LOF), transient overpower (TOP), and loss-of-heat-sink (LOHS) accidents

  14. Design basis and design features of WWER-440 model 213 nuclear power plants. Reference plant: Bohunice V2 (Slovakia)

    International Nuclear Information System (INIS)

    1994-05-01

    The prime objective of the IAEA Technical Co-operation Project on Evaluation of Safety Aspects of WWER-440 model 213 NPPs is to co-ordinate and to integrate assistance to national organizations in studying selected aspects of safety for the same type of reactors. Consequently, the study integrated the results generated by national activities carried out in the Czech Republic, Hungary, Slovakia and Ukraine and co-ordinated through the IAEA. Valuable assistance in carrying out the tasks was also provided by Bulgaria and Poland. A set of publications is being prepared to present the results of the project. The publications are intended to facilitate the review and utilization of the results of the project. They are also providing assistance in further refinement and/or extension of plant specific safety evaluation of model 213 NPPs. This Technical Document addressing the design basis and safety related design features of WWER-440 model 213 plants is the first of the series to be published. It is hoped that this document will be useful to anyone working in the field of WWER safety, and in particular to experts planning, executing or reviewing studies related to the subject. Refs, 36 figs, tabs

  15. GT-MHR design, performance, and safety

    International Nuclear Information System (INIS)

    Neylan, A.J.; Shenoy, A.; Silady, F.A.; Dunn, T.D.

    1994-11-01

    The Gas Turbine-Modular Helium Reactor (GT-MHR) is the result of coupling the evolution of a low power density passively safe modular reactor with key technology developments in the U.S. during the last decade: large industrial gas turbines; large active magnetic bearings; and compact, highly effective plate-fin heat exchangers. This is accomplished through the unique use of the Brayton cycle to produce electricity with the helium as primary coolant from the reactor directly driving the gas turbine electrical generator. This cycle can achieve a high net efficiency in the range of 45% to 48%. In the design of the GT-MHR the desirable inherent characteristics of the inert helium coolant, graphite core, and the coated fuel particles are supplemented with specific design features such as passive heat removal to achieve the safety objective of not disturbing the normal day-to-day activities of the public even for beyond design basis rare accidents. Each GT-MHR plant consists of four modules. The GT-MHR module components are contained within steel pressure vessels: a reactor vessel, a power conversion vessel, and a connecting cross vessel. All vessels are sited underground in a concrete silo, which serves as an independent vented low pressure containment structure. By capitalizing on industrial and aerospace gas turbine development, highly effective heat exchanger designs, and inherent gas cooled reactor temperature characteristics, the passively safe GT-MHR provides a sound technical, monetary, and environmental basis for new nuclear power generating capacity. This paper provides an update on the status of the design, which has been under development on the US-DOE program since February 1993. An assessment of plant performance and safety is also included

  16. Reactor safety under design basis flood condition for inland sites

    International Nuclear Information System (INIS)

    Hajela, S.; Bajaj, S.S.; Samota, A.; Verma, U.S.P.; Warudkar, A.S.

    2002-01-01

    Full text: In June 1994, there was an incident of flooding at Kakrapar Atomic Power Station (KAPS) due to combination of heavy rains and mechanical failure in the operation of gates at the adjoining weir. An indepth review of the incident was carried out and a number of flood protection measures were recommended and were implemented at site. As part of this review, a safety analysis was also done to demonstrate reactor safety with a series of failures considered in the flood protection features. For each inland NPP site, as part of design, different flood scenarios are analysed to arrive at design basis flood (DBF) level. This level is estimated based on worst combination of heavy local precipitation, flooding in river, failure of upstream/downstream water control structures

  17. Safety assessment in plant layout design using indexing approach: implementing inherent safety perspective. Part 2-Domino Hazard Index and case study.

    Science.gov (United States)

    Tugnoli, Alessandro; Khan, Faisal; Amyotte, Paul; Cozzani, Valerio

    2008-12-15

    The design of layout plans requires adequate assessment tools for the quantification of safety performance. The general focus of the present work is to introduce an inherent safety perspective at different points of the layout design process. In particular, index approaches for safety assessment and decision-making in the early stages of layout design are developed and discussed in this two-part contribution. Part 1 (accompanying paper) of the current work presents an integrated index approach for safety assessment of early plant layout. In the present paper (Part 2), an index for evaluation of the hazard related to the potential of domino effects is developed. The index considers the actual consequences of possible escalation scenarios and scores or ranks the subsequent accident propagation potential. The effects of inherent and passive protection measures are also assessed. The result is a rapid quantification of domino hazard potential that can provide substantial support for choices in the early stages of layout design. Additionally, a case study concerning selection among various layout options is presented and analyzed. The case study demonstrates the use and applicability of the indices developed in both parts of the current work and highlights the value of introducing inherent safety features early in layout design.

  18. Technical evaluation of the electrical, instrumentation, and control design aspects of the override of containment purge valve isolation and other engineered safety feature signals for the Fort Calhoun Nuclear Power Plant

    International Nuclear Information System (INIS)

    Hackett, D.B.

    1980-01-01

    This report documents the technical evaluation of the electrical, instrumentation, and control design aspects of the override of containment purge valve isolation and other engineered safety feature signals for the Fort Calhoun nuclear power plant. The review criteria are based on IEEE Std-279-1971 requirements for the safety signals to all purge and ventilation isolation valves. This report is supplied as part of the Selected Electrical, Instrumentation, and Control Systems Issues Program being conducted for the US Nuclear Regulatory Commission by Lawrence Livermore Laboratory

  19. Engineering design guidelines for nuclear criticality safety

    International Nuclear Information System (INIS)

    Waltz, W.R.

    1988-08-01

    This document provides general engineering design guidelines specific to nuclear criticality safety for a facility where the potential for a criticality accident exists. The guide is applicable to the design of new SRP/SRL facilities and to major modifications Of existing facilities. The document is intended an: A guide for persons actively engaged in the design process. A resource document for persons charged with design review for adequacy relative to criticality safety. A resource document for facility operating personnel. The guide defines six basic criticality safety design objectives and provides information to assist in accomplishing each objective. The guide in intended to supplement the design requirements relating to criticality safety contained in applicable Department of Energy (DOE) documents. The scope of the guide is limited to engineering design guidelines associated with criticality safety and does not include other areas of the design process, such as: criticality safety analytical methods and modeling, nor requirements for control of the design process

  20. Safety systems and features of boiling and pressurized water reactors

    International Nuclear Information System (INIS)

    Khair, H. O. M.

    2012-06-01

    The safe operation of nuclear power plants (NPP) requires a deep understanding of the functioning of physical processes and systems involved. This study was carried out to present an overview of the features of safety systems of boiling and pressurized water reactors that are available commercially. Brief description of purposes and functions of the various safety systems that are employed in these reactors was discussed and a brief comparison between the safety systems of BWRs and PWRs was made in an effort to emphasize of safety in NPPs.(Author)

  1. Safety culture in design. Final report

    International Nuclear Information System (INIS)

    Macchi, L.; Pietikaeinen, E.; Liinasuo, M.; Savioja, P.; Reiman, T.; Wahlstroem, M.; Kahlbom, U.; Rollenhagen, C.

    2013-04-01

    In this report we approach design from a safety culture approach As this research area is new and understudied, we take a wide scope on the issue. Different theoretical perspectives that can be taken when improving safety of the design process are considered in this report. We suggest that in the design context the concept of safety culture should be expanded from an organizational level to the level of the network of organizations involved in the design activity. The implication of approaching the design process from a safety culture perspective are discussed and the results of the empirical part of the research are presented. In the interview study in Finland and Sweden we identified challenges and opportunities in the design process from safety culture perspective. Also, a small part of the interview study concentrated on state of the art human factors engineering (HFE) practices in Finland and the results relating to that are presented. This report provide a basis for future development of systematic good design practices and for providing guidelines that can lead to safe and robust technical solutions. (Author)

  2. Safety culture in design. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Macchi, L.; Pietikaeinen, E.; Liinasuo, M.; Savioja, P.; Reiman, T.; Wahlstroem, M. [VTT Technical Research Centre of Finland, Espoo (Finland); Kahlbom, U. [Risk Pilot AB, Stockholm (Sweden); Rollenhagen, C. [Vattenfall, Stockholm, (Sweden)

    2013-04-15

    In this report we approach design from a safety culture approach As this research area is new and understudied, we take a wide scope on the issue. Different theoretical perspectives that can be taken when improving safety of the design process are considered in this report. We suggest that in the design context the concept of safety culture should be expanded from an organizational level to the level of the network of organizations involved in the design activity. The implication of approaching the design process from a safety culture perspective are discussed and the results of the empirical part of the research are presented. In the interview study in Finland and Sweden we identified challenges and opportunities in the design process from safety culture perspective. Also, a small part of the interview study concentrated on state of the art human factors engineering (HFE) practices in Finland and the results relating to that are presented. This report provide a basis for future development of systematic good design practices and for providing guidelines that can lead to safe and robust technical solutions. (Author)

  3. Seismic design and performance of nuclear safety related RC structures based on new seismic design principle

    International Nuclear Information System (INIS)

    Murugan, R.; Sivathanu Pillai, C.; Chattopadhyaya, S.; Sundaramurthy, C.

    2011-01-01

    Full text: Seismic design of safety related Reinforced Concrete (RC) structures of Nuclear power plants (NPP) in India as per the present AERB codal procedures tries to ensure predominantly elastic behaviour under OBE so that the features of Nuclear Power Plant (NPP) necessary for continued safe operation are designed to remain functional and prevent accident (collapse) of NPP under SSE for which certain Structures, Systems and Components (SSCs) those are necessary to ensure the capability to shut down the reactor safely, are designed to remain functional. While the seismic design principles of non safety related structures as per Indian code (IS 1893-2002) are ensuring elastic behaviour under DBE and inelastic behaviour under MCE by utilizing ductility and energy dissipation capacity of the structure effectively. The design principle of AERB code is ensuring elastic behaviour under OBE and is not enlightening much inference about the overall structural behaviour under SSE (only ensuring the capability of certain SSCs required for safe shutdown of reactor). Various buildings and structures of Indian Nuclear power plant are classified from the basis of associated safety functions in a descending order in according with their roles in preventions and mitigation of an accident or support functions for prevention. This paper covers a comprehensive seismic analysis and design methodology based on the AERB codal provisions followed for safety related RC structure taking Diesel Generator Building of PFBR as a case study and study and investigates its performance under OBE and SSE by carrying out Non-linear static Pushover analysis. Based on the analysis, observed variations, recommendations are given for getting the desired performance level so as to implement performance based design in the future NPP design

  4. Approach to design of future FBRs with enhanced safety and economy

    International Nuclear Information System (INIS)

    Raghupathy, S.; Balasubramaniyan, V.; Puthiyavinayagam, P.; Selvaraj, P.; Chellapandi, P.; Chetal, S.C.; Raj, Baldev

    2009-01-01

    After the techno economic demonstration of FBR technology through PFBR, it is essential to achieve high economic competitiveness with enhanced safety on par with other power generation options. Towards this, design studies were made with the objectives of identifying means to achieve enhanced safety, design features for improving economy, conceptualise the plant layout, ways to reduce construction time and to achieve higher capacity factor. This paper describes the outcome of the above studies and the approach to the design of future FBRs with enhanced safety and economy. High economic competitiveness and public acceptance are essential for commercial deployment of fast reactors. India has a well-laid out three phase nuclear power programme with Pressurised Heavy Water Reactors (PHWRs) utilizing natural uranium as fuel in the first phase and fast breeder reactors (FBRs) using plutonium and thorium in the second and third phases. Deploying FBRs on a commercial scale is vital for India in order to utilize the vast thorium reserves and to meet the long term energy needs of the country. As a first step, a 40 MWt, 13.5 MWe Fast Breeder Test Reactor (FBTR) was constructed and has been in operation since 1985. Operation of FBTR has given valuable operating experience with critical systems of FBRs including sodium systems and has given confidence to undertake the design of 500 MWe Prototype Fast Breeder Reactor (PFBR). Construction of PFBR is currently under progress and is expected to be completed by 2010. It is planned to construct four more oxide fuelled FBRs of 500 MWe capacity by 2020 and metal fuelled FBRs of 1000 MWe thereafter. Enhanced safety is proposed to be achieved through improvements in shut down systems, adopting in-vessel primary sodium purification concept, improving the reliability of decay heat removal system, shifting component decontamination activities to a separate building located outside the reactor containment building, providing additional

  5. Identifying Facilitators and Barriers for Patient Safety in a Medicine Label Design System Using Patient Simulation and Interviews

    DEFF Research Database (Denmark)

    Dieckmann, Peter; Clemmensen, Marianne Hald; Sørensen, Trine Kart

    2016-01-01

    Objectives Medicine label design plays an important role in improving patient safety. This study aimed at identifying facilitators and barriers in a medicine label system to prevent medication errors in clinical use by health care professionals. Methods The study design is qualitative and explora......Objectives Medicine label design plays an important role in improving patient safety. This study aimed at identifying facilitators and barriers in a medicine label system to prevent medication errors in clinical use by health care professionals. Methods The study design is qualitative...... of the system and some inconsistencies (different meaning of colors) posed challenges, when considered with the actual application context, in which there is little time to get familiar with the design features. Conclusions For optimizing medicine labels and obtaining the full benefit of label design features...

  6. Application of project management methodology in design management of nuclear safety related structure

    International Nuclear Information System (INIS)

    Chen Mao

    2004-01-01

    This paper focuses on the application of project management methodology in the design management of Nuclear Safety Related Structure (NSRS), considering the design management features of its civil construction. Based on the experiences from the management of several projects, the project management triangle is proposed to be used in the management, to well treat the position of design interface in the project management. Some other management methods are also proposed

  7. Cigarette Design Features: Effects on Emission Levels, User Perception, and Behavior.

    Science.gov (United States)

    Talhout, Reinskje; Richter, Patricia A; Stepanov, Irina; Watson, Christina V; Watson, Clifford H

    2018-01-01

    This paper describes the effects of non-tobacco, physical cigarette design features on smoke emissions, product appeal, and smoking behaviors - 3 factors that determine smoker's exposure and related health risks. We reviewed available evidence for the impact of filter ventilation, new filter types, and cigarettes dimensions on toxic emissions, smoker's perceptions, and behavior. For evidence sources we used scientific literature and websites providing product characteristics and marketing information. Whereas filter ventilation results in lower machine-generated emissions, it also leads to perceptions of lighter taste and relative safety in smokers who can unwittingly employ more intense smoking behavior to obtain the desired amount of nicotine and sensory appeal. Filter additives that modify smoke emissions can also modify sensory cues, resulting in changes in smoking behavior. Flavor capsules increase the cigarette's appeal and novelty, and lead to misperceptions of reduced harm. Slim cigarettes have lower yields of some smoke emissions, but smoking behavior can be more intense than with standard cigarettes. Physical design features significantly impact machine-measured emission yields in cigarette smoke, product appeal, smoking behaviors, and exposures in smokers. The influence of current and emerging design features is important in understanding the effectiveness of regulatory actions to reduce smoking-related harm.

  8. Balanced Design of Safety Systems of CAREM Advanced Reactor

    International Nuclear Information System (INIS)

    Grinblat, Pablo; Gimenez, Marcelo; Schlamp, Miguel

    2003-01-01

    Nuclear Power Plants must meet the performance that the market and the population demand in order to be part of the electricity supply industry.It is related mainly with the results of reactor's economy and safety.New advances in the methodology developed for reactor economic optimization analyzing its safety at an early engineering stage, aiming at balancing these important features of the design, are presented in this work.In particular, the coupling that appears when dimensioning the Emergency Injection System, the Residual Heat Removal System and the containment height of CAREM reactor is described.The new models appended to the computer code that embodies the methodology to balance de designs are shown.Finally the results obtained with the optimizations when applying it are presented.Furthermore, a criterion to establish the maximal diameter for acceptable breaks in RPV's penetrations arises from this work.The application of the methodology and the computer code developed turns out to prove the advantages they provide to reactor design so that the plants are properly balanced and optimized

  9. PIPEX - A model of a design concept for reprocessing plants with improved containment and surveillance features

    International Nuclear Information System (INIS)

    1979-03-01

    This paper explains that the PIPEX concept is essentially a reprocessing plant using the PUREX process but with in-built improved containment and surveillance features resulting in increased health protection and environmental safety as well as higher resistance to diversion of fissile material. The paper gives a general description of the design and operating philosophy of such a plant and goes on to examine the safeguards and safety principles and implications

  10. Engineering and safety features of modular vault dry storage

    International Nuclear Information System (INIS)

    Deacon, D.; Wheeler, D.J.

    1984-01-01

    This paper discusses the need for interim dry storage and reviews detailed features of the Modular Vault Dry storage concept. The concept meets three basic utility requirements. Firstly, the technology and safety features have been demonstrated on existing plant; secondly, it can be built and licensed in an acceptably short timescale; and thirdly, economic analysis shows that a modular vault dry store is often the cheapest option for interim storage

  11. Technical and institutional safety features of nuclear power plants in Brazil

    International Nuclear Information System (INIS)

    Rosa, L.P.

    1986-01-01

    This work reports technical, political and institutional safety features of nuclear power plants in Brazil. It is mainly concerned with reactor accidents and personnel safety. The three mile Island and Chernobyl accidents are also discussed and taken as examples. (A.C.A.S.)

  12. The verification methodologies for a software modeling of Engineered Safety Features- Component Control System (ESF-CCS)

    International Nuclear Information System (INIS)

    Lee, Young-Jun; Cheon, Se-Woo; Cha, Kyung-Ho; Park, Gee-Yong; Kwon, Kee-Choon

    2007-01-01

    The safety of a software is not guaranteed through a simple testing of the software. The testing reviews only the static functions of a software. The behavior, dynamic state of a software is not reviewed by a software testing. The Ariane5 rocket accident and the failure of the Virtual Case File Project are determined by a software fault. Although this software was tested thoroughly, the potential errors existed internally. There are a lot of methods to solve these problems. One of the methods is a formal methodology. It describes the software requirements as a formal specification during a software life cycle and verifies a specified design. This paper suggests the methods which verify the design to be described as a formal specification. We adapt these methods to the software of a ESF-CCS (Engineered Safety Features-Component Control System) and use the SCADE (Safety Critical Application Development Environment) tool for adopting the suggested verification methods

  13. Specific features of goal setting in road traffic safety

    Science.gov (United States)

    Kolesov, V. I.; Danilov, O. F.; Petrov, A. I.

    2017-10-01

    Road traffic safety (RTS) management is inherently a branch of cybernetics and therefore requires clear formalization of the task. The paper aims at identification of the specific features of goal setting in RTS management under the system approach. The paper presents the results of cybernetic modeling of the cause-to-effect mechanism of a road traffic accident (RTA); in here, the mechanism itself is viewed as a complex system. A designed management goal function is focused on minimizing the difficulty in achieving the target goal. Optimization of the target goal has been performed using the Lagrange principle. The created working algorithms have passed the soft testing. The key role of the obtained solution in the tactical and strategic RTS management is considered. The dynamics of the management effectiveness indicator has been analyzed based on the ten-year statistics for Russia.

  14. Education and Training of Safety Regulation for Nuclear Safety Infrastructure: Its Necessity and Unique Features

    International Nuclear Information System (INIS)

    Choi, Young Sung; Choi, Young Joon; Lee, Jae Cheon

    2009-01-01

    Faced with global warming and electricity demands, countries over the world recognize the comparative advantages of nuclear energy. It is estimated that about 300 nuclear power plants (NPPs) expect to be constructed until 2030 worldwide. In addition, according to the IAEA, approximately 20 new countries might have their first NPP in operation by 2030 in the high projection compared with bout 5 new countries in the low projection. When introducing nuclear power, the implementation of an appropriate infrastructure to address all of the relevant issues is a central concern of international community. In particular, nuclear power program requires, at an earlier stage than when construction starts, the development of a legal and regulatory framework and training of regulators and safety experts whose combined knowledge adequately covers all areas of nuclear safety and regulation applied at a NPP construction and operation. As an essential component of such human resource development, special attention was paid to the provision of education and training to regulators of which countries plan to introduce NPPs. In term of education theory, safety regulation has some unique features in learning and teaching, which are different from those of nuclear engineering or development. This paper overviews nuclear safety infrastructure, explores the roles of exporting countries, and presents features and components in education of nuclear safety regulation

  15. Design of control and safety rod and its drive mechanism of PFBR

    International Nuclear Information System (INIS)

    Rajan Babu, V.; Govindarajan, S.; Chetal, S.C.

    1997-01-01

    Control and Safety Rod (CSR) is one of the two types of absorber rods in shutdown systems of PFBR. Control and Safety Rod Drive Mechanism (CSRDM) actuates CSR to have vertical translatory motion in reactor core. The dual responsibilities entrusted on CSR to control reactor power during normal operating condition and to shutdown the reactor by scram action during abnormal condition, necessitate highly reliable design, analysis, testing and surveillance of CSR and CSRDM. The paper discusses on the salient features of CSR and CSRDM and design and analysis of individual sub-assemblies, viz., gripper, scram-release electromagnet, hydraulic dash pot, seals. Also it discusses on the developmental activities proposed and surveillance test requirements. (author)

  16. Perspective on Secure Development Activities and Features of Safety I and C Systems

    International Nuclear Information System (INIS)

    Kang, Youngdoo; Yu, Yeong Jin; Kim, Hyungtae; Kwon, Yong il; Park, Yeunsoo; Choo, Jaeyul; Son, Jun Young; Jeong, Choong Heui

    2015-01-01

    The Enforcement Decree of the Act on Physical Protection and Radiological Emergency (ED-APPRE) was revised December 2013 to include security requirements on computer systems at nuclear facilities to protect those systems against malicious cyber-attacks. It means Cyber-Security-related measures, controls and activities of safety I and C systems against cyber-attacks shall meet the requirements of ED-APPRE. Still regulation upon inadvertent access or non-malicious modifications to the safety I and C systems is covered under the Nuclear Safety Act. The objective of this paper is to propose KINS' regulatory perspective on secure development and features against non-malicious access or modification of safety I and C systems. Secure development activities and features aim to prevent inadvertent and non-malicious access, and to prevent unwanted action from personnel or connected systems for ensuring reliable operation of safety I and C systems. Secure development activities of safety I and C systems are life cycle activities to ensure unwanted, unneeded and undocumented code is not incorporated into the systems. Secure features shall be developed, verified and qualified throughout the development life cycle

  17. Perspective on Secure Development Activities and Features of Safety I and C Systems

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Youngdoo; Yu, Yeong Jin; Kim, Hyungtae; Kwon, Yong il; Park, Yeunsoo; Choo, Jaeyul; Son, Jun Young; Jeong, Choong Heui [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-05-15

    The Enforcement Decree of the Act on Physical Protection and Radiological Emergency (ED-APPRE) was revised December 2013 to include security requirements on computer systems at nuclear facilities to protect those systems against malicious cyber-attacks. It means Cyber-Security-related measures, controls and activities of safety I and C systems against cyber-attacks shall meet the requirements of ED-APPRE. Still regulation upon inadvertent access or non-malicious modifications to the safety I and C systems is covered under the Nuclear Safety Act. The objective of this paper is to propose KINS' regulatory perspective on secure development and features against non-malicious access or modification of safety I and C systems. Secure development activities and features aim to prevent inadvertent and non-malicious access, and to prevent unwanted action from personnel or connected systems for ensuring reliable operation of safety I and C systems. Secure development activities of safety I and C systems are life cycle activities to ensure unwanted, unneeded and undocumented code is not incorporated into the systems. Secure features shall be developed, verified and qualified throughout the development life cycle.

  18. Review of the Safety Design Approaches in Sodium Fast Reactors

    International Nuclear Information System (INIS)

    Suk, Soo Dong; Lee, Yong Bum

    2009-12-01

    The principle of the Defense in depth is essential in securing the safety of nuclear power plants, that is, to prevent cores-damaging severs accidents and to minimize the radiological consequences of the accidents 'as low as possible' (ALARA). One of the major design features of sodium fast reactors (SFRs) is that it has a large amount of sodium in the reactor vessel, providing a large heat capacity, such that it is feasible to contain the consequences of sever core damaging accidents in the vessel and primary system boundary. Containment of a severe accident in the primary system boundary, that is called in-vessel retention(IVR), is not a licensing requirement but set up as a design goal in most of the SFR design in the context of risk minimization. The objective of this report is to broadly review and compare the approaches and efforts made in the some of the major SFR designs of the US, Europe and Japan to prevent severe accidents and mitigate their consequences should they occur. Specifically, the subjects described in this report include design criteria or requirements, accident categorization and acceptance criteria, design features to prevent and contain severs accidents

  19. Investigation of ability to guess safety signs based on cognitive features in one of the petrochemical industries

    Directory of Open Access Journals (Sweden)

    G. A. Shirali

    2015-07-01

    .Conclusion: According to results of this study, use of principles of ergonomic design of signs and training are necessary to promote the ability to guess the safety signs to the minimum available standards. Therefore, it is possible to balance cognitive features especially “familiarity”, with the lowest score, and “meaningfulness” and “semantic closeness”, with the highest influential relationship with the ability to guess of signs. The developed regression model for this industry can be used to predict the ability to guess of safety signs in future studies

  20. Nuclear Electric flask design and safety case development during the last ten years

    International Nuclear Information System (INIS)

    Dougall, I.; Jones, D.K.

    1994-01-01

    The scope of Nuclear Electric's (NE) requirements for irradiated fuel transport and the relevant safety and regulatory compliance standards are outlined. To illustrate NE's approach to demonstrating package design compliance the basis of the safety case for unbottled Magnox fuel transport in the Mk M2 Magnox flask is described. The considerations which lay behind the development of the Mk A2 AGR Flask are indicated and a description is given of the main design features of this flask. Finally there is a summary of progress in obtaining Type B(M) approvals, based on the 1985 IAEA Regulations, for the requisite range of flask contents. (author)

  1. Mitigating construction safety risks using prevention through design.

    Science.gov (United States)

    Gangolells, Marta; Casals, Miquel; Forcada, Núria; Roca, Xavier; Fuertes, Alba

    2010-04-01

    Research and practice have demonstrated that decisions made prior to work at construction sites can influence construction worker safety. However, it has also been argued that most architects and design engineers possess neither the knowledge of construction safety nor the knowledge of construction processes necessary to effectively perform Construction Hazards Prevention through Design (CHPtD). This paper introduces a quantitative methodology that supports designers by providing a way to evaluate the safety-related performance of residential construction designs using a risk analysis-based approach. The methodology compares the overall safety risk level of various construction designs and ranks the significance of the various safety risks of each of these designs. The methodology also compares the absolute importance of a particular safety risk in various construction designs. Because the methodology identifies the relevance of each safety risk at a particular site prior to the construction stage, significant risks are highlighted in advance. Thus, a range of measures for mitigating safety risks can then be implemented during on-site construction. The methodology is specially worthwhile for designers, who can compare construction techniques and systems during the design phase and determine the corresponding level of safety risk without their creative talents being restricted. By using this methodology, construction companies can improve their on-site safety performance. Copyright 2010 Elsevier Ltd. All rights reserved.

  2. Safety design philosophy of Mitsubishi PWRs

    International Nuclear Information System (INIS)

    Hakata, T.; Kitamura, T.

    1993-01-01

    The basic safety design philosophy of Mitsubishi pressurized water reactors (PWRs) is discussed and compared with the British PWR. PWR plants are designed in accordance with the Japanese regulatory guidelines which are similar to American and International Atomic Energy Agency (IAEA) safety criteria and are based on defence-in-depth principles. The high reliability of nuclear power plants is especially emphasized in Mitsubishi PWRs, and this has been demonstrated by the good operating experience of PWR plants in Japan. The safety system designs of six key items, which were discussed in the recent review of overseas designs by British utilities, are addressed to show the difference in the design philosophy between the United Kingdom and Japan. (Author)

  3. Segmentation Scheme for Safety Enhancement of Engineered Safety Features Component Control System

    International Nuclear Information System (INIS)

    Lee, Sangseok; Sohn, Kwangyoung; Lee, Junku; Park, Geunok

    2013-01-01

    Common Caused Failure (CCF) or undetectable failure would adversely impact safety functions of ESF-CCS in the existing nuclear power plants. We propose the segmentation scheme to solve these problems. Main function assignment to segments in the proposed segmentation scheme is based on functional dependency and critical function success path by using the dependency depth matrix. The segment has functional independence and physical isolation. The segmentation structure is that prohibit failure propagation to others from undetectable failures. Therefore, the segmentation system structure has robustness to undetectable failures. The segmentation system structure has functional diversity. The specific function in the segment defected by CCF, the specific function could be maintained by diverse control function that assigned to other segments. Device level control signals and system level control signals are separated and also control signal and status signals are separated due to signal transmission paths are allocated independently based on signal type. In this kind of design, single device failure or failures on signal path in the channel couldn't result in the loss of all segmented functions simultaneously. Thus the proposed segmentation function is the design scheme that improves availability of safety functions. In conventional ESF-CCS, the single controller generates the signal to control the multiple safety functions, and the reliability is achieved by multiplication within the channel. This design has a drawback causing the loss of multiple functions due to the CCF (Common Cause Failure) and single failure Heterogeneous controller guarantees the diversity ensuring the execution of safety functions against the CCF and single failure, but requiring a lot of resources like manpower and cost. The segmentation technology based on the compartmentalization and functional diversification decreases the CCF and single failure nonetheless the identical types of controllers

  4. Segmentation Scheme for Safety Enhancement of Engineered Safety Features Component Control System

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sangseok; Sohn, Kwangyoung [Korea Reliability Technology and System, Daejeon (Korea, Republic of); Lee, Junku; Park, Geunok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    Common Caused Failure (CCF) or undetectable failure would adversely impact safety functions of ESF-CCS in the existing nuclear power plants. We propose the segmentation scheme to solve these problems. Main function assignment to segments in the proposed segmentation scheme is based on functional dependency and critical function success path by using the dependency depth matrix. The segment has functional independence and physical isolation. The segmentation structure is that prohibit failure propagation to others from undetectable failures. Therefore, the segmentation system structure has robustness to undetectable failures. The segmentation system structure has functional diversity. The specific function in the segment defected by CCF, the specific function could be maintained by diverse control function that assigned to other segments. Device level control signals and system level control signals are separated and also control signal and status signals are separated due to signal transmission paths are allocated independently based on signal type. In this kind of design, single device failure or failures on signal path in the channel couldn't result in the loss of all segmented functions simultaneously. Thus the proposed segmentation function is the design scheme that improves availability of safety functions. In conventional ESF-CCS, the single controller generates the signal to control the multiple safety functions, and the reliability is achieved by multiplication within the channel. This design has a drawback causing the loss of multiple functions due to the CCF (Common Cause Failure) and single failure Heterogeneous controller guarantees the diversity ensuring the execution of safety functions against the CCF and single failure, but requiring a lot of resources like manpower and cost. The segmentation technology based on the compartmentalization and functional diversification decreases the CCF and single failure nonetheless the identical types of

  5. Improvement of Safety Features in Standard Operation Procedure of Tc-99m Generator

    International Nuclear Information System (INIS)

    Manisah Saedon; Mohd Khairul Hakimi; Shyen, A.K.S.

    2011-01-01

    This paper describes the improvements proposed to the original production procedures for Tc-99m generators. Improvements are intended to add safety and health features for workers into the existing procedures. The difference between the new safe work procedures from the original work procedures; is the concern about the safety and health of employees other than the product safety. One of the suggested safety characteristics is by using the visual aid so that the workers can easily see and read the procedures when they perform their duties, whereas the previous procedures are kept in the manual and difficult to access. The purpose of this paper is to share information about the importance of safety and health features for the workers in the procedures established in addition to provide awareness to all parties involved. (author)

  6. Appendix C: safety design rationale

    International Nuclear Information System (INIS)

    Ghose, S.

    1985-01-01

    A brief discussion of the rationale for safety design of fusion plants is presented in the main text. Further detail safety considerations are presented in this appendix in the form of charts and tables. The author present some of the major safety criteria and other criteria used in blanket selection here

  7. Design features to reduce occupational exposure

    International Nuclear Information System (INIS)

    Adam, J.A.; DiSabatino, A.A. Jr.; Vanasse, R.E.

    1975-01-01

    Some of the design principles which are important considerations in the evolution of a nuclear power plant design to ensure that occupational radiation exposures can be minimized are discussed. Emphasis is placed on the design features affecting the basic layout and equipment locations within the plant. Examples are provided showing how these design objectives are realized in the Stone and Webster Reference Nuclear Power Plant Design, with particular emphasis on the Annulus Building portion of the reference plant. Design features which are useful in reducing occupational exposure during normal operation are discussed initially, followed by those that chiefly affect exposures during maintenance activity. Finally, those provisions in the design which assist in preventing the spread of radioactive contamination are presented

  8. Design safety improvements of Kozloduy NPP

    International Nuclear Information System (INIS)

    Hinovski, I.

    1999-01-01

    Design safety improvements of Kozloduy NPP, discussed in detail, are concerned with: primary circuit integrity; reactor pressure vessel integrity; primary coolant piping integrity; primary coolant overpressure protection; leak before break status; design basis accidents and transients; severe accident analysis; improvements of safety and support systems; containment/confinement leak tightness and strength; seismic safety improvements; WWER-1000 control rod insertion; upgrading and modernization of Units 5 and 6; Year 2000 problem

  9. Safety in the design of production lines

    DEFF Research Database (Denmark)

    Dyhrberg, Mette Bang; Broberg, Ole; Jacobsen, Peter

    2006-01-01

    This paper is a case study report on how safety considerations were handled in the process of redesigning a production line. The design process was characterized as a specification and negotiation process between engineers from the company and the supplier organization. The new production line...... in the specification material nor in their face-to-face meetings with the supplier. Safety aspects were not part of their work practice. On this basis, it was suggested that formal guidelines or procedures for integrating safety in the design of production lines would have no effect. Instead, the researchers set up...... became safer, but not as a result of any intentional plan to integrate safety aspects into the design process. Instead, the supplier’s design of a new piece of equipment had a higher built-in safety level. The engineering team in the company was aware of the importance of safety aspects neither...

  10. Preliminary safety design analysis of KALIMER

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Soo Dong; Kwon, Y. M.; Kim, K. D. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-03-01

    The national long-term R and D program updated in 1997 requires Korea Atomic Energy Research Institute(KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor (KALIMER), along with supporting R and D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 MWe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self consistent design meeting a set of the major safety design requirements for accident prevention. Some of current emphasis include those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve supporting R and D programs of substance. This document first introduces a set of safety design requirements and accident evaluation criteria established for the conceptual design of KALIMER and then summarizes some of the preliminary results of engineering and design analyses performed for the safety of KALIMER. 19 refs., 19 figs., 6 tabs. (Author)

  11. Seismic design features of the ACR Nuclear Power Plant

    International Nuclear Information System (INIS)

    Elgohary, M.; Saudy, A.; Aziz, T.

    2003-01-01

    Through their worldwide operating records, CANDU Nuclear Power Plants (NPPs) have repeatedly demonstrated safe, reliable and competitive performance. Currently, there are fourteen CANDU 6 single unit reactors operating or under construction worldwide. Atomic Energy of Canada Limited's (AECL) Advanced CANDU Reactor - the ACR. - is the genesis of a new generation of technologically advanced reactors founded on the CANDU reactor concept. The ACR is the next step in the evolution of the CANDU product line. The ACR products (ACR-700 and ACR-1000) are based on CANDU 6 (700 MWe class) and CANDU 9 (900 MWe class) reactors, therefore continuing AECL's successful approach of offering CANDU plants that appeal to a broad segment of the power generation market. The ACR products are based on the proven CANDU technology and incorporate advanced design technologies. The ACR NPP seismic design complies with Canadian standards that were specifically developed for nuclear seismic design and also with relevant International Atomic Energy Agency (IAEA) Safety Design Standards and Guides. However, since the ACR is also being offered to several markets with many potential sites and different regulatory environments, there is a need to develop a comprehensive approach for the seismic design input parameters. These input parameters are used in the design of the standard ACR product that is suitable for many sites while also maintaining its economic competitiveness. For this purpose, the ACR standard plant is conservatively qualified for a Design Basis Earthquake (DBE) with a peak horizontal ground acceleration of 0.3g for a wide range of soil/rock foundation conditions and Ground Response Spectra (GRS). These input parameters also address some of the current technical issues such as high frequency content and near field effects. In this paper, the ACR seismic design philosophy and seismic design approach for meeting the safety design requirements are reviewed. Also the seismic design

  12. Preliminary Analysis of a Steam Line Break Accident with the MARS-KS code for the SMART Design with Passive Safety Systems

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Doohyuk; Ko, Yungjoo; Suh, Jaeseung [Hannam Univ., Daejeon (Korea, Republic of); Bae, Hwang; Ryu, Sunguk; Yi, Sungjae; Park, Hyunsik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    SMART has been developed by KAERI, and SMART-Standard Design Approval (SDA) was recently granted in 2012. A SMART design with Passive Safety System (PSS) features (called SMART-PSS) is being developed and added to the standard design of SMART by KAERI to improve its safety system. Active safety systems such as safety injection pumps will be replaced by a passive safety system, which is actuated only by the gravity force caused by the height difference. All tanks for the passive safety systems are higher than the injection nozzle, which is located around the reactor coolant pumps (RCPs). In this study, a preliminary analysis of the main steam line break accident (MSLB) was performed using the MARS-KS code to understand the general behavior of the SMART-PSS design and to prepare its validation test with the SMART-ITL (FESTA) facility. An anticipated accident for the main steam line break (MSLB) was performed using the MARS-KS code to understand the thermal-hydraulic behaviors of the SMART-PSS design. The preliminary analysis provides good insight into the passive safety system design features of the SMART-PSS and the thermal-hydraulic characteristics of the SMART design. The analysis results of the MSLB showed that the core water collapsed level inside the core support barrel was maintained high over the active core top level during the transient period. Therefore, the SMART-PSS design has satisfied the requirements to maintain the plant at a safe shutdown condition during 72 hours without AC power or operator action after an anticipated accident.

  13. Fire Safety Design of Wood Structures

    DEFF Research Database (Denmark)

    Hertz, Kristian Dahl

    2006-01-01

    Lecture Notes on Fire Safety Design of Wood Structures including charring of wood and load bearing capacity of beams, columns, and connections.......Lecture Notes on Fire Safety Design of Wood Structures including charring of wood and load bearing capacity of beams, columns, and connections....

  14. General design safety principles for nuclear power plants

    International Nuclear Information System (INIS)

    1986-01-01

    This Safety Guide provides the safety principles and the approach that have been used to implement the Code in the Safety Guides. These safety principles and the approach are tied closely to the safety analyses needed to assist the design process, and are used to verify the adequacy of nuclear power plant designs. This Guide also provides a framework for the use of other design Safety Guides. However, although it explains the principles on which the other Safety Guides are based, the requirements for specific applications of these principles are mostly found in the other Guides

  15. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (French Edition)

    International Nuclear Information System (INIS)

    2017-01-01

    This publication establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication.

  16. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Russian Edition)

    International Nuclear Information System (INIS)

    2016-01-01

    This publication establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication.

  17. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Arabic Edition)

    International Nuclear Information System (INIS)

    2017-01-01

    This publication establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication.

  18. Nuclear safety cooperation for Soviet designed reactors

    International Nuclear Information System (INIS)

    Reisman, A.W.; Horak, W.C.

    1995-01-01

    The nuclear accident at the Chernobyl nuclear power plant in 1986 first alerted the West to the significant safety risks of Soviet designed reactors. Five years later, this concern was reaffirmed when the IAEA, as a result of a review by an international team of nuclear safety experts, announced that it did not believe the Kozloduy nuclear power plants in Bulgaria could be operated safely. To address these safety concerns, the G-7 summit in Munich in July 1992 outlined a five point program to address the safety problems of Soviet Designed Reactors: operational safety improvement; near-term technical improvements to plants based on safety assessment; enhancing regulatory regimes; examination of the scope for replacing less safe plants by the development of alternative energy sources and the more efficient use of energy; and upgrading of the plants of more recent design. As of early 1994, over 20 countries and international organizations have pledged hundreds of millions of dollars in financial assistance to improve safety. This paper summarizes these assistance efforts for Soviet designed reactors, draws lessons learned from these activities, and offers some options for better addressing these concerns

  19. Evaluation of features to support safety and quality in general practice clinical software

    Science.gov (United States)

    2011-01-01

    Background Electronic prescribing is now the norm in many countries. We wished to find out if clinical software systems used by general practitioners in Australia include features (functional capabilities and other characteristics) that facilitate improved patient safety and care, with a focus on quality use of medicines. Methods Seven clinical software systems used in general practice were evaluated. Fifty software features that were previously rated as likely to have a high impact on safety and/or quality of care in general practice were tested and are reported here. Results The range of results for the implementation of 50 features across the 7 clinical software systems was as follows: 17-31 features (34-62%) were fully implemented, 9-13 (18-26%) partially implemented, and 9-20 (18-40%) not implemented. Key findings included: Access to evidence based drug and therapeutic information was limited. Decision support for prescribing was available but varied markedly between systems. During prescribing there was potential for medicine mis-selection in some systems, and linking a medicine with its indication was optional. The definition of 'current medicines' versus 'past medicines' was not always clear. There were limited resources for patients, and some medicines lists for patients were suboptimal. Results were provided to the software vendors, who were keen to improve their systems. Conclusions The clinical systems tested lack some of the features expected to support patient safety and quality of care. Standards and certification for clinical software would ensure that safety features are present and that there is a minimum level of clinical functionality that clinicians could expect to find in any system.

  20. The role of probabilistic safety assessment in the design

    International Nuclear Information System (INIS)

    Green, A.; Ingham, E.L.

    1989-01-01

    The use of probabilistic safety assessment (PSA) for Heysham 2 and Torness marked a major change in the design approach to nuclear safety within the U.K. Design Safety Guidelines incorporating probabilistic safety targets required that design justification would necessitate explicit consideration of the consequence of accidents in relation to their frequency. The paper discusses these safety targets and their implications, the integration of PSA into the design process and an outline of the methodology. The influence of PSA on the design is discussed together with its role in the overall demonstration of reactor safety. (author)

  1. Review on JMTR safety design for LEU core conversion

    International Nuclear Information System (INIS)

    Komori, Yoshihiro; Yokokawa, Makoto; Saruta, Toru; Inada, Seiji; Sakurai, Fumio; Yamamoto, Katsumune; Oyamada, Rokuro; Saito, Minoru

    1993-12-01

    Safety of the JMTR was fully reviewed for the core conversion to low enriched uranium fuel. Fundamental policies for the JMTR safety design were reconsidered based on the examination guide for safety design of test and research reactors, and safety of the JMTR was confirmed. This report describes the safety design of the JMTR from the viewpoint of major functions for reactor safety. (author)

  2. Safety requirements applicable to the SMART design

    International Nuclear Information System (INIS)

    Seul, Kwang Won; Kim, Wee Kyong; Kim, Hho Jung

    1999-01-01

    The 330 MW thermal power of integral reactor, named SMART (System integrated Modular Advanced ReacTor), is under development at KAERI for seawater desalination application and electricity generation. The final product of nuclear desalination plant (NDP) is electricity and fresh water. Thus, in addition to the protection of the public around the plant facility from the possible release of radioactive materials, the fresh water should be prevented from radioactivity contamination. In this study, to ensure the safety of SMART reactor in the early stage of design development, the safety requirements applicable to the SMART design were investigated, based on the current regulatory requirements for the existing NPPs and the advanced light water reactor (LWR) designs. The interface requirements related to the desalination facility were also investigated, based on the recent IAEA research activities pertaining to the NDP. As a result, it was found that the current regulatory requirements and guidance for the existing NPPs and advanced LWR designs are applicable to the SMART design and its safety evaluation. However, the safety requirements related to the SMART-specific design and the desalination plant are needed to develop in the future to assure the safety of the SMART reactor

  3. Design of diverse safety rod and its drive mechanism of PFBR

    International Nuclear Information System (INIS)

    Vijayashree, R.; Govindarajan, S.; Chetal, S.C.

    1997-01-01

    In Prototype Fast Breeder Reactor (PFBR), there are two types of absorber rods for control and shutdown of the reactor in the event of any abnormal event. They are: (i) Control and Safety Rod (CSR) and (ii) the Diverse Safety Rod (DSR). Of these, the former (CSR) caters to the control function of the reactor during normal operating conditions and to the shutdown during abnormal situations. The DSR, on the other hand is meant essentially for the reactor shutdown to take care of any abnormal transient. It is rather important to note that functionally the DSR is independent of CSR in the sense, that it can bring the reactor to a cold shutdown state and maintain it even under the hypothetical condition of the failure of CSR. From the design point of view, this stipulates a failure probability of less than 10 -4 per demand. The DSR is normally parked above the core by the Diverse Safety Rod Drive Mechanism (DSRDM). On receiving a scram signal it gets released from the holding electromagnet and falls under the gravity into the core. Diverse features are incorporated both in the absorber rods and in the drive mechanisms to avoid common mode failures. This paper discusses the salient features of DSR and DSRDM. A brief account of detailed design, analysis and development of two important subassemblies viz. electromagnet and sodium dash pot is also presented. In addition, a brief comparison between CSR and DSR including their drive mechanisms is also provided. (author)

  4. Safety research needs for Russian-designed reactors. Requirements situation

    International Nuclear Information System (INIS)

    Brown, R. Allan; Holmstrom, Heikki; Reocreux, Michel; Schulz, Helmut; Liesch, Klaus; Santarossa, Giampiero; Hayamizu, Yoshitaka; Asmolov, Vladimir; Bolshov, Leonid; Strizhov, Valerii; Bougaenko, Sergei; Nikitin, Yuri N.; Proklov, Vladimir; Potapov, Alexandre; Kinnersly, Stephen R.; Voronin, Leonid M.; Honekamp, John R.; Frescura, Gianni M.; Maki, Nobuo; Reig, Javier; ); Bekjord, Eric S.; Rosinger, Herbert E.

    1998-01-01

    In June 1995, an OECD Support Group was set up to perform a broad study of the safety research needs of Russian-designed reactors. The emphasis of the study is on the VVER-type reactors in part because of the larger base of knowledge within the NEA Member countries related to LWRs. For the RBMKs, the study does not make the judgement that such reactors can be brought to acceptable levels of safety but focuses on near term efforts that can contribute to reducing the risk to the public. The need for the safety research must be evaluated in context of the lifetime of the reactors. The principal outcome of the work of the Support Group is the identification of a number of research topics which the members believe should receive priority attention over the next several years if risk levels are to be reduced and public safety enhanced. These appear in the Conclusions and Recommendations section of the report, and are the following: - The most important near-term need for VVER and RBMK safety research is to establish a sound technical basis for the emergency operating procedures used by the plant staff to prevent or halt the progression of accidents (i.e., Accident Management) and for plant safety improvements. - Co-operation of Western and Eastern experts should help to avoid East-West know-how gaps in the future, as safety technology continues to improve. - Safety research in Eastern countries will make an important contribution to public safety as it has in OECD countries. - RBMK safety research, including verification of codes, starts from a smaller base of experience than VVER, and is at an earlier stage of development. Technical Conclusions: - Research to improve human performance and operational safety of VVER and RBMK plants is extremely important. - VVER thermal-hydraulic and reactor physics research should focus on full validation of codes to VVER-specific features, and on extension of experimental data base. - Methods of assessing VVER pressure boundary

  5. Specialists' meeting on passive and active safety features of LMFRs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1991-07-01

    The objective of the meeting was to discuss and exchange information on passive and active safety concepts and to find some reasonable coupling of these concept, aiming at firmer establishment of plant safety and at the same time of plant cost reduction. The following main topical areas were discussed by delegates: (1) Overview - review of national status on the safety design approaches of LMFRs (2) Safety characteristics of decay heat removal system (DHRS) (3) Safety characteristics of reactor protection system (RPS) and reactor shutdown system (RSS) (4) Core safety characteristics.

  6. Specialists' meeting on passive and active safety features of LMFRs

    International Nuclear Information System (INIS)

    1991-01-01

    The objective of the meeting was to discuss and exchange information on passive and active safety concepts and to find some reasonable coupling of these concept, aiming at firmer establishment of plant safety and at the same time of plant cost reduction. The following main topical areas were discussed by delegates: (1) Overview - review of national status on the safety design approaches of LMFRs (2) Safety characteristics of decay heat removal system (DHRS) (3) Safety characteristics of reactor protection system (RPS) and reactor shutdown system (RSS) (4) Core safety characteristics

  7. An object-oriented feature-based design system face-based detection of feature interactions

    International Nuclear Information System (INIS)

    Ariffin Abdul Razak

    1999-01-01

    This paper presents an object-oriented, feature-based design system which supports the integration of design and manufacture by ensuring that part descriptions fully account for any feature interactions. Manufacturing information is extracted from the feature descriptions in the form of volumes and Tool Access Directions, TADs. When features interact, both volumes and TADs are updated. This methodology has been demonstrated by developing a prototype system in which ACIS attributes are used to record feature information within the data structure of the solid model. The system implemented in the C++ programming language and embedded in a menu-driven X-windows user interface to the ACIS 3D Toolkit. (author)

  8. Safety design guides for fire protection for CANDU 9

    International Nuclear Information System (INIS)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; A. C. D. Wright

    1996-03-01

    This safety design guide establishes design requirements to ensure the radiological risk to the public due to fire is acceptable and operating personnel are adequately protected from the hazards of fires. This safety design guide also specifies the safety criteria for fire protection to be applied to mitigate fires and recommends the fire protection program to be established to initiate, coordinate and document the design activities associated with fire protection. The requirements for fire protection outlined in this safety design guide shall be satisfied in the design stage and the change status of the regulatory requirements, code and standards should be traced and incorporated into this safety design guide accordingly. 1 fig., (Author) .new

  9. IAEA safety standards and approach to safety of advanced reactors

    International Nuclear Information System (INIS)

    Gasparini, M.

    2004-01-01

    The paper presents an overview of the IAEA safety standards including their overall structure and purpose. A detailed presentation is devoted to the general approach to safety that is embodied in the current safety requirements for the design of nuclear power plants. A safety approach is proposed for the future. This approach can be used as reference for a safe design, for safety assessment and for the preparation of the safety requirements. The method proposes an integration of deterministic and risk informed concepts in the general frame of a generalized concept of safety goals and defence in depth. This methodology may provide a useful tool for the preparation of safety requirements for the design and operation of any kind of reactor including small and medium sized reactors with innovative safety features.(author)

  10. Designing attractive gamification features for collaborative storytelling websites.

    Science.gov (United States)

    Hsu, Shang Hwa; Chang, Jen-Wei; Lee, Chun-Chia

    2013-06-01

    Gamification design is considered as the predictor of collaborative storytelling websites' success. Although aforementioned studies have mentioned a broad range of factors that may influence gamification, they neither depicted the actual design features nor relative attractiveness among them. This study aims to identify attractive gamification features for collaborative storytelling websites. We first constructed a hierarchical system structure of gamification design of collaborative storytelling websites and conducted a focus group interview with eighteen frequent users to identify 35gamification features. After that, this study determined the relative attractiveness of these gamification features by administrating an online survey to 6333 collaborative storytelling websites users. The results indicated that the top 10 most attractive gamification features could account for more than 50% of attractiveness among these 35 gamification features. The feature of unpredictable time pressure is important to website users, yet not revealed in previous relevant studies. Implications of the findings were discussed.

  11. Safety design guide for safety related systems for CANDU 9

    International Nuclear Information System (INIS)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; A. C. D. Wright

    1996-03-01

    In general, two types of safety related systems and structures exist in the nuclear plant; The one is a systems and structures which perform safety functions during the normal operation of the plant, and the other is a systems and structures which perform safety functions to mitigate events caused by failure of the normally operating systems or by naturally occurring phenomena. In this safety design guide, these systems are identified in detail, and the major events for which the safety functions are required and the major safety requirements are identified in the list. As the probabilistic safety assessments are completed during the course of the project, additions or deletions to the list may be justified. 3 tabs. (Author) .new

  12. Safety design guide for safety related systems for CANDU 9

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young [Korea Atomic Energy Research Institute, Daeduk (Korea, Republic of); Wright, A.C.D. [Atomic Energy of Canada Ltd., Toronto (Canada)

    1996-03-01

    In general, two types of safety related systems and structures exist in the nuclear plant; The one is a systems and structures which perform safety functions during the normal operation of the plant, and the other is a systems and structures which perform safety functions to mitigate events caused by failure of the normally operating systems or by naturally occurring phenomena. In this safety design guide, these systems are identified in detail, and the major events for which the safety functions are required and the major safety requirements are identified in the list. As the probabilistic safety assessments are completed during the course of the project, additions or deletions to the list may be justified. 3 tabs. (Author) .new.

  13. Design features of Advanced Power Reactor (APR) 1400 steam generator

    International Nuclear Information System (INIS)

    Park, Tae-Jung; Park, Jun-Soo; Kim, Moo-Yong

    2004-01-01

    Advanced Power Reactor 1400 (APR 1400) which is to achieve the improvement of the safety and economical efficiency has been developed by Korea Hydro and Nuclear Power Co., Ltd. (KHNP) with the support from industries and research institutes. The steam generator for APR 1400 is an evolutionary type from System 80 + , which is the recirculating U-tube heat exchanger with integral economizer. Compared to the System 80 + steam generator, it is focused on the improved design features, operating and design conditions of APR 1400 steam generator. Especially, from the operation experience of Korean Standard Nuclear Power Plant (KSNP) steam generator, the lessons-learned measures are incorporated to prevent the tube wear caused by flow-induced vibration (FIV). The concepts for the preventive design features against FIV are categorized to two fields; flow distribution and dynamic response characteristics. From the standpoint of flow distribution characteristics, the egg-crate flow distribution plate (EFDP) is installed to prevent the local excessive flow loaded on the most susceptible tube to wear. The parametric study is performed to select the optimum design with the efficient mitigation of local excessive flow. ATHOS3 Mod-01 is used and partly modified to analyze the flow field of the APR 1400 steam generator. In addition, the upper tube bundle support is designed to eliminate the presence of tube with a low natural frequency. Based on the improved upper tube bundle support, the modal analysis is performed and compared with that of System 80 + . Using the results of flow distribution and modal analysis, the two mechanisms of flow-induced vibration are investigated; fluid-elastic instability (FEI) and random turbulence excitation (RTE). (authors)

  14. OSHA and Experimental Safety Design.

    Science.gov (United States)

    Sichak, Stephen, Jr.

    1983-01-01

    Suggests that a governmental agency, most likely Occupational Safety and Health Administration (OSHA) be considered in the safety design stage of any experiment. Focusing on OSHA's role, discusses such topics as occupational health hazards of toxic chemicals in laboratories, occupational exposure to benzene, and role/regulations of other agencies.…

  15. Cascade Distillation System Design for Safety and Mission Assurance

    Science.gov (United States)

    Sarguisingh, Miriam; Callahan, Michael R.; Okon, Shira

    2015-01-01

    Per the NASA Human Health, Life Support and Habitation System Technology Area 06 report "crewed missions venturing beyond Low-Earth Orbit (LEO) will require technologies with improved reliability, reduced mass, self-sufficiency, and minimal logistical needs as an emergency or quick-return option will not be feasible".1 To meet this need, the development team of the second generation Cascade Distillation System (CDS 2.0) chose a development approach that explicitly incorporate consideration of safety, mission assurance, and autonomy. The CDS 2.0 preliminary design focused on establishing a functional baseline that meets the CDS core capabilities and performance. The critical design phase is now focused on incorporating features through a deliberative process of establishing the systems failure modes and effects, identifying mitigation strategies, and evaluating the merit of the proposed actions through analysis and test. This paper details results of this effort on the CDS 2.0 design.

  16. Innovative safety ideas for fusion experimental machines

    International Nuclear Information System (INIS)

    Brereton, S.J.; Gouge, M.; Piet, S.J.; Merrill, B.J.; Holland, D.F.; Sze, D.K.

    1990-01-01

    Throughout the early stages of design of fusion experimental devices, such as ITER, safety experts have worked with designers to incorporate safety features into the design. Recent efforts have focused on passive safety features. Although designs of near-term fusion machines may appear consistent with expected regulatory requirements, the safety characteristics can potentially be more attractive. Here, a variety of suggestions that appear promising in terms of improving safety are given. These include new concepts, innovative technologies, further support of past concepts, and possible modification to operating scenarios. Some technical discussion on the feasibility of the proposals is provided. The ideas are generally conceptual at this stage and require further assessment and development work. However, each has the potential for enhancing the safety of experimental devices. 33 refs., 6 figs., 9 tabs

  17. A Distributed Feature-based Environment for Collaborative Design

    Directory of Open Access Journals (Sweden)

    Wei-Dong Li

    2003-02-01

    Full Text Available This paper presents a client/server design environment based on 3D feature-based modelling and Java technologies to enable design information to be shared efficiently among members within a design team. In this environment, design tasks and clients are organised through working sessions generated and maintained by a collaborative server. The information from an individual design client during a design process is updated and broadcast to other clients in the same session through an event-driven and call-back mechanism. The downstream manufacturing analysis modules can be wrapped as agents and plugged into the open environment to support the design activities. At the server side, a feature-feature relationship is established and maintained to filter the varied information of a working part, so as to facilitate efficient information update during the design process.

  18. Design of a Construction Safety Training System using Contextual Design Methodology

    OpenAIRE

    Baldev, Darshan H.

    2006-01-01

    In the U.S., the majority of construction companies are small companies with 10 or fewer employees (BLS, 2004). The fatality rate in the construction industry is high, indicating a need for implementing safety training to a greater extent. This research addresses two main goals: to make recommendations and design a safety training system for small construction companies, and to use Contextual Design to design the training system. Contextual Design was developed by Holtzblatt (Beyer and Holtzb...

  19. 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

  20. Design and qualification of HPD based designs for safety systems

    International Nuclear Information System (INIS)

    Sharma, Mukesh Kr.; Chavan, Madhavi A.; Sawhney, Pratibha A.; Mohanty, Ashutos; John, Ajith K.; Ganesh, G.

    2014-01-01

    Field Programmable Gate Arrays (FPGA) and Complex Programmable Logic Devices (CPLD) are increasingly being used in C and I system of NPPs. The function of such an integrated circuit is not defined by the supplier of the physical component or micro-electronic technology but by the C and I designer. The hardware subsystems implemented in these devices typically use Hardware Description Language (HDL) like VHDL or Verilog to describe the functionality at the design entry level. These circuits are commonly known as 'HDL-Programmed Devices', (HPD). RCnD has developed a set of hardware boards to be used in next generation C and I systems. The boards have been designed based on present day technology and components. The intelligence of these boards has been implemented in HPDs (FPGA/CPLD) using VHDL. Since these boards are used in the safety and safety related systems, they have undergone a rigorous V and V process and qualification tests. This paper discusses the design attributes and qualification of these HPD based designs for nuclear class safety systems. (author)

  1. NUCLEAR SAFETY DESIGN BASES FOR LICENSE APPLICATION

    International Nuclear Information System (INIS)

    Garrett, R.J.

    2005-01-01

    The purpose of this report is to identify and document the nuclear safety design requirements that are specific to structures, systems, and components (SSCs) of the repository that are important to safety (ITS) during the preclosure period and to support the preclosure safety analysis and the license application for the high-level radioactive waste (HLW) repository at Yucca Mountain, Nevada. The scope of this report includes the assignment of nuclear safety design requirements to SSCs that are ITS and does not include the assignment of design requirements to SSCs or natural or engineered barriers that are important to waste isolation (ITWI). These requirements are used as input for the design of the SSCs that are ITS such that the preclosure performance objectives of 10 CFR 63.111 [DIRS 156605] are met. The natural or engineered barriers that are important to meeting the postclosure performance objectives of 10 CFR 63.113 [DIRS 156605] are identified as ITWI. Although a structure, system, or component (SSC) that is ITS may also be ITWI, this report is only concerned with providing the nuclear safety requirements for SSCs that are ITS to prevent or mitigate event sequences during the repository preclosure period

  2. Safety studies on Korean fusion DEMO plant using integrated safety assessment methodology

    International Nuclear Information System (INIS)

    Oh, Kyemin; Kang, Myoung-suk; Heo, Gyunyoung; Kim, Hyoung-chan

    2014-01-01

    Highlights: •The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant. •The concepts of integrated safety assessment methodology (ISAM) that can be applied in addressing regulatory requirements and recognizing safety issues for K-DEMO were emphasized. •Phenomena identification and ranking table (PIRT) was proposed. It can recognize vulnerabilities of systems and identify the gaps in technical areas requiring additional researches. •This work is expected to contribute on the conceptual design of safety features for K-DEMO to design engineers and the guidance for regulatory requirements to licensers. -- Abstract: The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant (K-DEMO) as a part of R and D program through the National Fusion Research Institute of Korea. Even though nuclear regulation and licensing framework is well setup due to the operating and design experience of Pressurized Water Reactors (PWRs) since 1970s, the regulatory authority of South Korea has concerns on the challenge of facing new nuclear facilities including K-DEMO due to the differences in systems, materials, and inherent safety feature from conventional PWRs. Even though the follow-up of the ITER license process facilitates to deal with significant safety issues of fusion facilities, a licensee as well as a licenser should identify the gaps between ITER and DEMO in terms of safety issues. First we reviewed the methods of conducting safety analysis for unprecedented nuclear facilities such as Generation IV reactors, particularly very high temperature reactor (VHTR), which is called as integrated safety assessment methodology (ISAM). Second, the analysis for the conceptual design of K-DEMO on the basis of ISAM was conducted. The ISAM consists of five analytical tools to develop the safety requirements from licensee

  3. Safety studies on Korean fusion DEMO plant using integrated safety assessment methodology

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Kyemin; Kang, Myoung-suk [Kyung Hee University, Youngin-si, Gyeonggi-do 446-701 (Korea, Republic of); Heo, Gyunyoung, E-mail: gheo@khu.ac.kr [Kyung Hee University, Youngin-si, Gyeonggi-do 446-701 (Korea, Republic of); Kim, Hyoung-chan [National Fusion Research Institute, Daejeon-si 305-333 (Korea, Republic of)

    2014-10-15

    Highlights: •The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant. •The concepts of integrated safety assessment methodology (ISAM) that can be applied in addressing regulatory requirements and recognizing safety issues for K-DEMO were emphasized. •Phenomena identification and ranking table (PIRT) was proposed. It can recognize vulnerabilities of systems and identify the gaps in technical areas requiring additional researches. •This work is expected to contribute on the conceptual design of safety features for K-DEMO to design engineers and the guidance for regulatory requirements to licensers. -- Abstract: The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant (K-DEMO) as a part of R and D program through the National Fusion Research Institute of Korea. Even though nuclear regulation and licensing framework is well setup due to the operating and design experience of Pressurized Water Reactors (PWRs) since 1970s, the regulatory authority of South Korea has concerns on the challenge of facing new nuclear facilities including K-DEMO due to the differences in systems, materials, and inherent safety feature from conventional PWRs. Even though the follow-up of the ITER license process facilitates to deal with significant safety issues of fusion facilities, a licensee as well as a licenser should identify the gaps between ITER and DEMO in terms of safety issues. First we reviewed the methods of conducting safety analysis for unprecedented nuclear facilities such as Generation IV reactors, particularly very high temperature reactor (VHTR), which is called as integrated safety assessment methodology (ISAM). Second, the analysis for the conceptual design of K-DEMO on the basis of ISAM was conducted. The ISAM consists of five analytical tools to develop the safety requirements from licensee

  4. Design aspects of radiological safety in nuclear facilities

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  5. Mechanical Design Features of PGSFR NSSS

    International Nuclear Information System (INIS)

    Park, Chang-Gyu; Koo, Gyeong-Hoi; Cho, Jae-Hun; Kim, Sung-Kyun

    2016-01-01

    The NSSS(Nuclear Steam Supply System) is composed of PHTS(Primary Heat Transport System), IHTS (Intermediate Heat Transport System), and SGS(Steam Generation System). And, DHRS(Decay Heat Removal System) adopts both the active and passive systems for diversity. The structures including components and piping should be designed to ensure the structural integrity for their design life against mechanical and operational loads. In this study, the mechanical design features for the structures and components that make up PGSFR NSSS are described. The mechanical design features of structures and components for a PGSFR NSSS are described. The structures are being designed to maintain the structural integrity for their design lifetime by considering the high temperature operating condition. The decay heat removal system(DHRS) removes all reactor decay heat in two ways; active type(ADHRS) and passive type(PDHRS). ADHRS consists of DHX, blower, FHX, circulation pump, and expansion tank. But PDHRS consists of DHX, AHX, and expansion tank. FHX is a finned-tube-type sodium-to-air heat exchanger whereas AHX is a helical-type sodium-to-air heat exchanger

  6. Mechanical Design Features of PGSFR NSSS

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chang-Gyu; Koo, Gyeong-Hoi; Cho, Jae-Hun; Kim, Sung-Kyun [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The NSSS(Nuclear Steam Supply System) is composed of PHTS(Primary Heat Transport System), IHTS (Intermediate Heat Transport System), and SGS(Steam Generation System). And, DHRS(Decay Heat Removal System) adopts both the active and passive systems for diversity. The structures including components and piping should be designed to ensure the structural integrity for their design life against mechanical and operational loads. In this study, the mechanical design features for the structures and components that make up PGSFR NSSS are described. The mechanical design features of structures and components for a PGSFR NSSS are described. The structures are being designed to maintain the structural integrity for their design lifetime by considering the high temperature operating condition. The decay heat removal system(DHRS) removes all reactor decay heat in two ways; active type(ADHRS) and passive type(PDHRS). ADHRS consists of DHX, blower, FHX, circulation pump, and expansion tank. But PDHRS consists of DHX, AHX, and expansion tank. FHX is a finned-tube-type sodium-to-air heat exchanger whereas AHX is a helical-type sodium-to-air heat exchanger.

  7. Relevant safety issues in designing the HTR-10 reactor

    International Nuclear Information System (INIS)

    Sun Yuliang; Xu Yuanghui

    2001-01-01

    The HTR-10 is a 10 MWth pebble bed high temperature gas cooled reactor being constructed as a research facility at the Institute of Nuclear Energy Technology. This paper discusses design issues of the HTR-10 which are related to safety. It addresses the safety criteria used in the development and assessment of the design, the safety important systems, and the safety classification of components. It also summarises the results of safety analysis, including the approach used for the radioactive source term, as well as the approach to containment design. (author)

  8. Safety and environmental requirements and design targets for TIBER-II

    International Nuclear Information System (INIS)

    Piet, S.J.

    1987-09-01

    A consistent set of safety and environmental requirements and design targets was proposed and adopted for the TIBER-II (Tokamak Ignition/Burn Experimental Reactor) design effort. TIBER-II is the most recent US version of a fusion experimental test reactor (ETR). These safety and environmental design targets were one contribution of the Fusion Safety Program in the TIBER-II design effort. The other contribution, safety analyses, is documented in the TIBER-II design report. The TIBER-II approach, described here, concentrated on logical development of, first, a complete and consistent set of safety and environmental requirements that are likely appropriate for an ETR, and, second, an initial set of design targets to guide TIBER-II. Because of limited time in the TIBER-II design effort, the iterative process only included one iteration - one set of targets and one design. Future ETR design efforts should therefore build on these design targets and the associated safety analyses. 29 refs., 5 figs., 3 tabs

  9. Safety methodology implementation in the conceptual design phase of a fusion reactor

    International Nuclear Information System (INIS)

    Rodriguez-Rodrigo, L.; Elbez-Uzan, J.

    2007-01-01

    The licensing of ITER in France represents the first process for licensing a fusion facility in the framework of an experimental device with a total Tritium inventory of 3 kg. The main ITER parameters are far from those expected in the future demonstration reactors where the fusion power will be at least 5 times higher and the additional heating power could also reach up to 5 times the one foreseen in ITER. Main safety requirements for these reactors are based, among other conditions, on their inherent features as low amount of fuel, very low impurity content of structural materials, minimum waste repository, no active systems for safe shut-down, and no need for evacuation of population after the most severe accident. The design of such reactors is at the stage of conceptual studies and is mainly dealing with plasma performances, tritium breeding, blanket/divertor designs and solution of engineering issues, as well as bounding accidents or classification of waste. The methodological approach for integrating safety analysis as a tool for optimizing the design of the overall fusion installation for future reactors in the conceptual design phase is sketched, including the machine itself and the different auxiliary nuclear buildings. (author)

  10. Safety features of TR-2 reactor

    International Nuclear Information System (INIS)

    Tuerker, T.

    2001-01-01

    TR-2 is a swimming pool type research reactor with 5 MW thermal power and uses standard MTR plate type fuel elements. Each standard fuel element consist of 23 fuel plates with a meat + cladding thickness of 0.127 cm, coolant channel clearance is 0.21 cm. Originally TR-2 is designed for %93 enriched U-Al. Alloy fuel meat.This work is based on the preparation of the Final Safety Analyses Report (FSAR) of the TR-2 reactor. The main aspect is to investigate the behaviour of TR-2 reactor under the accident and abnormal operating conditions, which cowers the accident spectrum unique for the TR-2 reactor. This presentation covers some selected transient analyses which are important for the safety aspects of the TR-2 reactor like reactivity induced startup accidents, pump coast down (Loss of Flow Accident, LOFA) and other accidents which are charecteristic to the TR-2

  11. Environmental, health, and safety by design

    International Nuclear Information System (INIS)

    Soklow, R.G.

    1999-01-01

    Solar Turbines Incorporated created a self-directed work team, the Safety and Environmental Awareness (SEA) Team that initiated a company wide effort to raise employee awareness to promote integrating responsible environmental, health, and safety practices into product design, manufacturing, and services. Environmental, health, and safety issues influence how all businesses operate around the world. Companies choose to operate in an environmentally responsible manner because it not only benefits employees and the communities where they live, it also benefits the business when superior performance results in a competitive advantage. Solar surveyed gas turbines users to identify their top environmental and safety concerns and issues. The authors asked about various environmental and safety aspects of their equipment. Results from the survey has helped engineering and design focus efforts so that future products and product improvements assist customers in meeting their regulatory obligations and social responsibilities. Air pollution has historically been one of the most important environmental issues facing customers, because pollutant emissions greatly influence equipment choices and operation flexibility. There are other environmental, health and safety issues: sustainable fire suppression choices, start systems, hazardous materials use and ability to recycle materials, package accessibility, noise and product take back issues

  12. Conceptual design study for the demonstration reactor of JSFR. (3) Safety design and evaluation

    International Nuclear Information System (INIS)

    Tani, Akihiro; Shimakawa, Yoshio; Kubo, Shigenobu; Fujimura, Ken; Yamano, Hidemasa

    2011-01-01

    This paper describes the result of conceptual safety design and evaluation for the demonstration plant of Japan sodium-cooled fast reactor (JSFR), which was preliminarily conducted for providing information necessary to decide the plant specification for further design study. The plant major specifications except for output power and safety design concept are almost the same as those of the commercial JSFR. A set of safety evaluation for typical design basis events (DBEs) is mainly focused here, which was conducted for the 750 MWe design. Safety analyses for DBEs evaluation were performed on the basis of conservative assumptions using a one-dimensional flow network code with point kinetics. For representative DBEs, transient over power type events and loss of flow type events were analyzed. The long-term loss-of-offsite power event was also calculated to evaluate the natural circulation decay heat removal system. All analytical results showed to meet tentative safety criteria, thus it was confirmed that the safety design concept of JSFR is feasible against DBEs. (author)

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

    International Nuclear Information System (INIS)

    1996-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-04-01

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

  15. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Russian Edition); Bezopasnost' atomnykh ehlektrostantsij: proektirovanie. Konkretnye trebovaniya bezopasnosti

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-04-15

    This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  16. Application of probabilistic methods to safety R and D and design choices

    International Nuclear Information System (INIS)

    Gavigan, F.X.; Griffith, J.D.

    1977-01-01

    The Liquid Metal Fast Breeder Reactor (LMFBR) safety program is committed to identifying and exploiting areas in which probabilistic methods can be developed and used in making reactor safety R and D choices and optimizing designs of safety systems. Emphasis will be placed on a positive approach of solidifying and expanding our knowledge. This will provide the groundwork for a consensus on FBR risk. The management structure which will be used is based on a mechanistic approach to an LMFBR Core Disruptive Accident (CDA) with risk partitioned into ''Lines of Assurance,'' i.e., independent, phenomenologically-based barriers which will impede or mitigate the progression and consequences of accident sequences. Quantitative determination of the probability of breach of these barriers through the completion of work identified for each Line of Assurance will allow the quantification of the contribution to risk reduction associated with the success of each barrier. This process can lead to better use of resources by channeling R and D in directions which promise the greatest potential for reducing risk and by identifying an orderly approach to the development and demonstration of design features which will keep LMFBR risks at an acceptable level

  17. Multinational Design Evaluation Programme (MDEP) - Safety Goals

    International Nuclear Information System (INIS)

    Vaughan, G.J.

    2011-01-01

    One of the aims of the NEA's Multinational Design Evaluation Programme (MDEP) is to work towards greater harmonisation of regulatory requirements. To achieve this aim, it is necessary that there is a degree of convergence on the safety goals that are required to be met by designers and operators. The term 'safety goals' is defined to cover all health and safety requirements which must be met: these may be deterministic rules and/or probabilistic targets. They should cover the safety of workers, public and the environment in line with the IAEA's Basic Safety Objective; encompassing safety in normal operation through to severe accidents. MDEP is also interested in how its work can be extended to future reactors, which may use significantly different technology to the almost ubiquitous LWRs used today and in the next generation, building on the close co-operation within MDEP between the regulators who are currently engaged in constructing or carrying out design reviews on new designs. For two designs this work has involved several regulators sharing their safety assessments and in some cases issuing statements on issues that need to be addressed. Work is also progressing towards joint regulatory position statements on specific assessment areas. Harmonisation of safety goals will enhance the cooperation between regulators as further developments in design and technology occur. All regulators have safety goals, but these are expressed in many different ways and exercises in comparing them frequently are done at a very low level eg specific temperatures in the reactor vessel of a specific reactor type. The differences in the requirements from different regulators are difficult to resolve as the goals are derived using different principles and assumptions and are often for a specific technology. Therefore a different approach is being investigated, starting with the top-level safety goals and try to derive a structure and means of deriving lower tier

  18. Nuclear Safety Design Base for License Application

    International Nuclear Information System (INIS)

    R.J. Garrett

    2005-01-01

    The purpose of this report is to identify and document the nuclear safety design requirements that are specific to structures, systems, and components (SSCs) of the repository that are important to safety (ITS) during the preclosure period and to support the preclosure safety analysis and the license application for the high-level radioactive waste (HLW) repository at Yucca Mountain, Nevada. The scope of this report includes the assignment of nuclear safety design requirements to SSCs that are ITS and does not include the assignment of design requirements to SSCs or natural or engineered barriers that are important to waste isolation (ITWI). These requirements are used as input for the design of the SSCs that are ITS such that the preclosure performance objectives of 10 CFR 63.111(b) [DIRS 173273] are met. The natural or engineered barriers that are important to meeting the postclosure performance objectives of 10 CFR 63.113(b) and (c) [DIRS 173273] are identified as ITWI. Although a structure, system, or component (SSC) that is ITS may also be ITWI, this report is only concerned with providing the nuclear safety requirements for SSCs that are ITS to prevent or mitigate event sequences during the repository preclosure period

  19. Working Towards Unified Safety Design Criteria for Modular High Temperature Gas-cooled Reactor Designs

    International Nuclear Information System (INIS)

    Reitsma, Frederik; Silady, Fred; Kunitomi, Kazuhiko

    2014-01-01

    The Nuclear Power Development Section of the IAEA recently received approval for a Coordinated Research Project (CRP) to investigate and make proposals on modular High Temperature Gas-cooled Reactor (HTGR) Safety design criteria. It is expected that these criteria would consider past experience and existing safety standards in the light of modular HTGR material and design characteristics to propose safety design criteria. It will consider the deterministic and risk-informed safety design standards that apply to the wide spectrum of Off- normal events under development worldwide for existing and planned HTGRs. The CRP would also take into account lessons from the Fukushima Daiichi accident, clarifying the safety approach and safety evaluation criteria for design and beyond design basis events, including those events that can affect multiple reactor modules and/or are dependent on the application proximate to the plant site. (e. g., industrial process steam/heat). The logical flow of criteria is from the fundamental inherent safety characteristics of modular HTGRs and associated expected performance characteristics, to the safety functions required to ensure those characteristics during the wide spectrum of Off-normal events, and finally to specific criteria related to those functions. This is detailed in the paper with specific examples included of how it may be applied. The results of the CRP will be made available to the member states and HTGR community. (author)

  20. Cold Vacuum Drying Safety Class Instrumentation and Control System Design Description

    International Nuclear Information System (INIS)

    WHITEHURST, R.

    1999-01-01

    This document describes the Cold Vacuum Drying Facility (CVDF) Safety Class Instrumentation and Control system (SCIC). The SCIC provides safety functions and features to protect the environment, off-site and on-site personnel and equipment. The function of the SCIC is to provide automatic trip features, valve interlocks, alarms, indication and control for the cold vacuum drying process

  1. Assistance for design and realization of real-time safety critical applications according to Oasis

    International Nuclear Information System (INIS)

    Aussagues, Ch.; Cordonnier, Ch.; Quetueil, I.; David, V.

    1998-01-01

    Assistance for design and realization is all the more essential when safety-critical, real-time and complex applications are considered. When developing such applications, real-time and dependability features should be managed as soon as possible. Then, since the design phase, programmers have to strive even more than with common applications. In the context of the OASIS approach, assistance for design and realization is based on some intrinsic properties of the model, i.e. determinism, behavior independence and timeliness. The assistance, that may be furnished in the OASIS approach, has three complementary components that constitute the focus of this article. (authors)

  2. Design aspects of safety critical instrumentation of nuclear installations

    Energy Technology Data Exchange (ETDEWEB)

    Swaminathan, P. [Electronics Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India)]. E-mail: swamy@igcar.ernet.in

    2005-07-01

    Safety critical instrumentation systems ensure safe shutdown/configuration of the nuclear installation when process status exceeds the safety threshold limits. Design requirements for safety critical instrumentation such as functional and electrical independence, fail-safe design, and architecture to ensure the specified unsafe failure rate and safe failure rate, human machine interface (HMI), etc., are explained with examples. Different fault tolerant architectures like 1/2, 2/2, 2/3 hot stand-by are compared for safety critical instrumentation. For embedded systems, software quality assurance is detailed both during design phase and O and M phase. Different software development models such as waterfall model and spiral model are explained with examples. The error distribution in embedded system is detailed. The usage of formal method is outlined to reduce the specification error. The guidelines for coding of application software are outlined. The interface problems of safety critical instrumentation with sensors, actuators, other computer systems, etc., are detailed with examples. Testability and maintainability shall be taken into account during design phase. Online diagnostics for safety critical instrumentation is detailed with examples. Salient details of design guides from Atomic Energy Regulatory Board, International Atomic Energy Agency and standards from IEEE, BIS are given towards the design of safety critical instrumentation systems. (author)

  3. Design aspects of safety critical instrumentation of nuclear installations

    International Nuclear Information System (INIS)

    Swaminathan, P.

    2005-01-01

    Safety critical instrumentation systems ensure safe shutdown/configuration of the nuclear installation when process status exceeds the safety threshold limits. Design requirements for safety critical instrumentation such as functional and electrical independence, fail-safe design, and architecture to ensure the specified unsafe failure rate and safe failure rate, human machine interface (HMI), etc., are explained with examples. Different fault tolerant architectures like 1/2, 2/2, 2/3 hot stand-by are compared for safety critical instrumentation. For embedded systems, software quality assurance is detailed both during design phase and O and M phase. Different software development models such as waterfall model and spiral model are explained with examples. The error distribution in embedded system is detailed. The usage of formal method is outlined to reduce the specification error. The guidelines for coding of application software are outlined. The interface problems of safety critical instrumentation with sensors, actuators, other computer systems, etc., are detailed with examples. Testability and maintainability shall be taken into account during design phase. Online diagnostics for safety critical instrumentation is detailed with examples. Salient details of design guides from Atomic Energy Regulatory Board, International Atomic Energy Agency and standards from IEEE, BIS are given towards the design of safety critical instrumentation systems. (author)

  4. Overview of Indian position: Passive and active safety features of LMFBRs

    International Nuclear Information System (INIS)

    Paranjpe, S.R.

    1991-01-01

    Since PWR and BWR type reactors are considered to have acceptable level of safety, there is a general statement that LMFBR type reactors should be designed to be 'at least as safe' as PWRs and BWRs. Nuclear safety issues considered include the following: flexibility of operation over the entire power range, effectiveness of steps ensuring the safety as well as plant availability under foreseeable incident conditions; guaranteed safe shut-down of the reactor; ability to remove safely and reliably the decay heat under shut-down conditions; containment capability including protection of the reactor from external events. It is concluded that liquid metal fast breeder reactors (LMFBRs) can be designed very easily to eliminate anxieties about their safety and no difficulties are expected in the licensing procedure of this type of reactors

  5. Safety and design limits

    International Nuclear Information System (INIS)

    Shishkov, L. K.; Gorbaev, V. A.; Tsyganov, S. V.

    2007-01-01

    The paper touches upon the issues of NPP safety ensuring at the stage of fuel load design and operation by applying special limitations for a series of parameters, that is, design limits. Two following approaches are compared: the one used by west specialists for the PWR reactor and the Russian approach employed for the WWER reactor. The closeness of approaches is established, differences that are mainly peculiarities of terms are noted (Authors)

  6. 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.

  7. Main physics features driving design concept and physics design constraints

    International Nuclear Information System (INIS)

    Fujisawa, Noboru; Sugihara, Masayoshi; Yamamoto, Shin

    1987-07-01

    Major physics design philosophies are described, which are essential bases for a plasma design and may have significant impacts on a reactor design concept. Those design philosophies are classified into two groups, physics design drivers and physics design constraints. The design drivers are featured by the fact that a designer is free to choose and the choice may be guided by his opinion, such as ignition, a pulse length, an operation scenario, etc.. The design constraints may follow a physical law, such as plasma confinement, β-limit, density limit, and so on. (author)

  8. Passive safety optimization in liquid-sodium cooled reactors

    International Nuclear Information System (INIS)

    Cahalan, J. E.; Hahn, D.; Chang, W.-P.; Kwon, Y.-M.; Nuclear Engineering Division; Korea Atomic Energy Research Inst.

    2004-01-01

    This report summarizes the results of a three-year collaboration between Argonne National Laboratory (ANL) and the Korea Atomic Energy Research Institute (KAERI) to identify and quantify the performance of innovative design features in metallic-fueled, sodium-cooled fast reactor designs. The objective of the work was to establish the reliability and safety margin enhancements provided by design innovations offering significant potential for construction, maintenance, and operating cost reductions. The project goal was accomplished with a combination of advanced model development (Task 1), analysis of innovative design and safety features (Tasks 2 and 3), and planning of key safety experiments (Task 4)

  9. 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.

  10. Design for safety: A cognitive engineering approach to the control and management of nuclear power plants

    International Nuclear Information System (INIS)

    Boy, Guy A.; Schmitt, Kara A.

    2013-01-01

    Highlights: ► Complexity must be understood and handled well in order to design for safety. ► Complexity can be reduced during design by using the AUTOS pyramid model. ► Procedures are human automation, much as software is machine automation. ► Identifying emergent behaviors reduces procedure accumulation. ► Human-in-the-loop-simulations help to understand emergent behaviors. -- Abstract: This paper presents an analytical approach to design for safety that is based on 30 years of experience in the field of Human-centered design. This field is often qualified as governing safety–critical systems where risk management is a crucial issue. We need to better understand what the main facets of safety are that should be taken into account during the design and development processes. There are many factors that contribute to design for safety. We propose some of these factors and an articulation of them from requirement gathering and synthesis to formative evaluations to summative evaluations. Among these factors, we analyze complexity, flexibility, stability, redundancy, support, training, experience and testing. However, we cannot design a safe and reliable product in one shot; design is incremental. A product and its various uses become progressively mature. When we deal with new products, issues come from the fact that practice features emerge from the use of the product and are difficult, even impossible, to predict ahead of time. The automation within is an important portion of this maturity, and must be understood well. This is why design for safety is not possible without anticipatory simulations and a period of tests in the real world, such as operational testing in nuclear power plants. In addition, designing for safety is not finished when the product is delivered; experience feedback, or human-in-the-loop simulation (HITLS) is an important part of the overall global design process. The AUTOS pyramid approach can assist in simplifying the

  11. 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

  12. Panel 1: Safety design criteria

    International Nuclear Information System (INIS)

    Yllera, Javier

    2013-01-01

    There is general consensus in the nuclear community, and more after the Fukushima accident, that the deployment of nuclear energy has to be done at the highest levels of nuclear safety and that safety cannot be compromised by other factors. It is well understood that reactors that are being licensed and the new generations of reactors that will be constructed in the future will need to reach higher safety levels than the existing ones. Several countries and international organizations or international groups are launching initiatives to harmonise safety goals, safety requirements, safety objectives, regulations, criteria or safety reference levels. There are differences in the meanings of these terms and the working approaches, but the overall purpose is the same: to specify how new plants can be safer. In this context, the IAEA has an statutory function for developing international nuclear safety standards. The IAEA safety standards are per se not mandatory for IAEA Member States. Regulating safety is a national responsibility, and many States have decided to adopt the IAEA’s standards for use in their national regulations in different ways. The IAEA Safety Standards represent international consensus on what must constitute a high level of safety for nuclear installations. In the area of NPP design, IAEA safety standards that are published are intended to apply primarily to new plants. It might not be practicable to apply all the requirements to plants that are already in operation. In addition, the focus is primarily on plants with water cooled reactors

  13. Design lessons from using programmable controllers in the MFTF-B personnel safety and interlocks system

    International Nuclear Information System (INIS)

    Branum, J.D.

    1983-01-01

    Applying programmable controllers in critical applications such as personnel safety and interlocks systems requires special considerations in the design of both hardware and software. All modern programmable controller systems feature extensive internal diagnostic capabilities to protect against problems such as program memory errors; however most, if not all present designs lack an intrinsic capability for detecting and countering failures on the field-side of their I/O modules. Many of the most common styles of I/O modules can also introduce potentially dangerous sneak circuits, even without component failure. This paper presents the most significant lessons learned to date in the design of the MFTF-B Personnel Safety and Interlocks System, which utilizes two non-redundant programmable controllers with over 800 I/O points each. Specific problems recognized during the design process as well as those discovered during initial testing and operation are discussed along with their specific solutions in hardware and software

  14. Safety research needs for Russian-designed reactors

    International Nuclear Information System (INIS)

    1998-01-01

    In June 1995, an OECD Support Group was set up to perform a broad study of the safety research needs of Russian-designed reactors. This Support Group was endorsed by the CSNI. The Support Group, which is composed of senior experts on safety research from several OECD countries and from Russia, prepared this Report. The Group reviewed the safety research performed to support Russian-designed reactors and set down its views on future needs. The review concentrates on the following main topics: Thermal-Hydraulics/Plant Transients for VVERs; Integrity of Equipment and Structures for VVERs; Severe Accidents for VVERs; Operational Safety Issues; Thermal-Hydraulics/Plant Transients for RBMKs; Integrity of Equipment and Structures for RBMKs; Severe Accidents for RBMKs. (K.A.)

  15. Design features of the KSTAR in-vessel control coils

    Energy Technology Data Exchange (ETDEWEB)

    Kim, H.K. [National Fusion Research Institute (NFRI), 52 Yeoeun-dong, Yusung-ku, Daejeon, 305-333 (Korea, Republic of)], E-mail: hkkim@nfri.re.kr; Yang, H.L.; Kim, G.H.; Kim, Jin-Yong; Jhang, Hogun; Bak, J.S.; Lee, G.S. [National Fusion Research Institute (NFRI), 52 Yeoeun-dong, Yusung-ku, Daejeon, 305-333 (Korea, Republic of)

    2009-06-15

    In-vessel control coils (IVCCs) are to be used for the fast plasma position control, field error correction (FEC), and resistive wall mode (RWM) stabilization for the Korea Superconducting Tokamak Advanced Research (KSTAR) device. The IVCC system comprises 16 segments to be unified into a single set to achieve following remarkable engineering advantages; (1) enhancement of the coil system reliability with no welding or brazing works inside the vacuum vessel, (2) simplification in fabrication and installation owing to coils being fabricated outside the vacuum vessel and installed after device assembly, and (3) easy repair and maintenance of the coil system. Each segment is designed in 8 turns coil of 32 mm x 15 mm rectangular oxygen free high conductive copper with a 7 mm diameter internal coolant hole. The conductors are enclosed in 2 mm thick Inconel 625 rectangular welded vacuum jacket with epoxy/glass insulation. Structural analyses were implemented to evaluate structural safety against electromagnetic loads acting on the IVCC for the various operation scenarios using finite element analysis. This paper describes the design features and structural analysis results of the KSTAR in-vessel control coils.

  16. PSA in design of passive/active safety reactors

    International Nuclear Information System (INIS)

    Sato, T.; Tanabe, A.; Kondo, S.

    1995-01-01

    PSAs in the design of advanced reactors are applied mainly in level 1 PSA areas. However, even in level 1 PSA, there are certain areas where special care must be taken depending on plant design concepts. This paper identifies these areas both for passive and active safety reactor concepts. For example, 'long-term PSA' and shutdown PSA are very important for a passive safety reactor concept from the standpoint of effectiveness of a grace period and passive safety systems. External events are also important for an active safety reactor concept. These kinds of special PSAs are difficult to conduct precisely in a conceptual design stage. This paper shows methods of conducting these kinds of special PSAs simply and conveniently and the use of acquired insights for the design of advanced reactors. This paper also clarifies the meaning or definition of a grace period from the standpoint of PSA

  17. Safety design integrated in the building delivery system

    DEFF Research Database (Denmark)

    Jørgensen, Kirsten

    2013-01-01

    . The purpose of this article is to demonstrate how safety and health can be integrated in the design phases integrated in the management delivery systems within construction, The method for the research was to go through the building delivery system step by step and create a normative description of what, when......In construction, it is important to view safety and health as an integrated part of the way that “designers” are working. The designers cowers architects, constructors, engineers and others who carry out their consulting services in the design phase of a construction project. The philosophy...... and how to fully integrate safety in each part of the process. The result is a concept and guideline including control forms for how to integrate safety design in the Building Delivery System plus what to do and when. The concept has been tested in an educational context. The practical value...

  18. LABORATORY DESIGN CONSIDERATIONS FOR SAFETY.

    Science.gov (United States)

    National Safety Council, Chicago, IL. Campus Safety Association.

    THIS SET OF CONSIDERATIONS HAS BEEN PREPARED TO PROVIDE PERSONS WORKING ON THE DESIGN OF NEW OR REMODELED LABORATORY FACILITIES WITH A SUITABLE REFERENCE GUIDE TO DESIGN SAFETY. THERE IS NO DISTINCTION BETWEEN TYPES OF LABORATORY AND THE EMPHASIS IS ON GIVING GUIDES AND ALTERNATIVES RATHER THAN DETAILED SPECIFICATIONS. AREAS COVERED INCLUDE--(1)…

  19. Design trade-offs in view of safety considerations

    International Nuclear Information System (INIS)

    Saji, G.; Kishida, K.; Inoue, T.

    1978-01-01

    In view of resolving conflicting demands of cost, safety, flexibility of operation and design margins, safety design of various plant systems is discussed referring to their weight on construction costs. An influence of hypothetical core disruptive accident (HCDA) and loss of piping integrity (LOPI) on plant design and thus on construction materials is discussed, in optimising future commercial FBR plants. (author)

  20. The enhancement of Ignalina NPP in design and operational safety

    International Nuclear Information System (INIS)

    Negrivoda, G.

    1999-01-01

    Enhancement of Ignalina NPP design include: core design improvements; fuel channel integrity (multiple pressure tube rupture); improvements of shutdown systems; improvements of instrumentation and control devices; containment strength and tightness; design basis accident analysis; improvements of safety and support systems; seismic safety enhancement; Year 2000 project; cracks in pipes. Enhancement of operational safety includes: quality assurance; configuration management; safety management and safety culture; emergency operating procedures; training and full scope simulator; in-service inspection; fire protection and ageing monitoring and management

  1. Main engineering features driving design concept and engineering design constraints

    International Nuclear Information System (INIS)

    Saito, Ryusei; Kobayashi, Takeshi; Yamada, Masao

    1987-09-01

    Major engineering design philosophies are described, which are essential bases for an engineering design and may have significant impacts on a reactor design concept. Those design philosophies are classified into two groups, engineering design drivers and engineering design constraints. The design drivers are featured by the fact that a designer is free to choose and the choice may be guided by his opinion, such as coil system, a mechanical configuration, a tritium breeding scenario, etc.. The design constraints may follow a natural law or engineering limit, such as material strength, coil current density, and so on. (author)

  2. Status of conceptual safety design study of Japanese sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Kubo, Shigenobu; Kurisaka, Kenichi; Niwa, Hajime; Shimakawa, Yoshio

    2005-01-01

    In this paper, the current conceptual safety design and related evaluation of Japanese Sodium-cooled Fast Reactor which is studied in the framework of the Feasibility Study (FS) on commercialized Fast Reactor Cycle Systems in Japan are described. The purpose of the safety design is to establish a feasible safety concept of FBR which aims at a sustainable energy source of the next generations. The safety targets and the safety design principle are set aiming at realizing worldwide acceptability of the safety level. The basic safety design concept, which can meet the safety targets, was formulated taking along with the defense-in-depth philosophy as the basic safety design principle. In order to cope with wide range of energy and resource demands, there are some various designs both of oxide and metal fuel for JSFR. Some analytical results of typical design basis events, design extension conditions and core damage frequency estimation show the feasibility of the safety design concept for them. (author)

  3. Evaluation of design feature No.20 -- Ground support options

    International Nuclear Information System (INIS)

    Duan, F.

    2000-01-01

    Ground support options are primarily evaluated for emplacement drifts while ground support systems for non-emplacement openings such as access mains and ventilation drifts are not evaluated against LADS evaluation criteria in this report. Considerations include functional requirements for ground support, the use of a steel-lined system, and the feasibility of using an unlined ground support system principally with grouted rock bolts for permanent ground support. The feature evaluation also emphasizes the postclosure effects of ground support materials on waste isolation and the preclosure aspects such as durability, maintainability, constructibility, safety, engineering acceptability, and cost. This evaluation is to: (A) Review the existing analyses, reports, and studies regarding this design feature, and compile relevant information on performance characteristics. (B) Develop an appropriate evaluation approach for evaluating ground support options against evaluation criteria provided by the LADS team. (C) Evaluate ground support options not only for their preclosure performance in terms of drift stability, material durability, maintenance, constructibility, and cost, but also for their postclosure performance in terms of chemical effects of ground support materials (i.e., concrete, steel) on waste isolation and radionuclide transport. Specifically, the scope for ground support options evaluation include: (1) all steel-lined drifts (no cementitious materials), (2) unlined drifts with minimum cementitious materials (e.g., grout for rockbolts), and (3) concrete-lined drifts, with the focus on the postclosure acceptability evaluation. In addition, unlined drifts with zero cementitious materials (e.g., use of frictional bolts such as split sets, Swellex bolts) are briefly discussed. (D) Identify candidate ground support systems that have the potential to enhance the repository performance based on the feature evaluation. and (E) Provide conclusions and recommendations

  4. Preliminary Safety Design Report for Remote Handled Low-Level Waste Disposal Facility

    Energy Technology Data Exchange (ETDEWEB)

    Timothy Solack; Carol Mason

    2012-03-01

    A new onsite, remote-handled low-level waste disposal facility has been identified as the highest ranked alternative for providing continued, uninterrupted remote-handled low-level waste disposal for remote-handled low-level waste from the Idaho National Laboratory and for nuclear fuel processing activities at the Naval Reactors Facility. Historically, this type of waste has been disposed of at the Radioactive Waste Management Complex. Disposal of remote-handled low-level waste in concrete disposal vaults at the Radioactive Waste Management Complex will continue until the facility is full or until it must be closed in preparation for final remediation of the Subsurface Disposal Area (approximately at the end of Fiscal Year 2017). This preliminary safety design report supports the design of a proposed onsite remote-handled low-level waste disposal facility by providing an initial nuclear facility hazard categorization, by discussing site characteristics that impact accident analysis, by providing the facility and process information necessary to support the hazard analysis, by identifying and evaluating potential hazards for processes associated with onsite handling and disposal of remote-handled low-level waste, and by discussing the need for safety features that will become part of the facility design.

  5. Design an optimum safety policy for personnel safety management - A system dynamic approach

    International Nuclear Information System (INIS)

    Balaji, P.

    2014-01-01

    Personnel safety management (PSM) ensures that employee's work conditions are healthy and safe by various proactive and reactive approaches. Nowadays it is a complex phenomenon because of increasing dynamic nature of organisations which results in an increase of accidents. An important part of accident prevention is to understand the existing system properly and make safety strategies for that system. System dynamics modelling appears to be an appropriate methodology to explore and make strategy for PSM. Many system dynamics models of industrial systems have been built entirely for specific host firms. This thesis illustrates an alternative approach. The generic system dynamics model of Personnel safety management was developed and tested in a host firm. The model was undergone various structural, behavioural and policy tests. The utility and effectiveness of model was further explored through modelling a safety scenario. In order to create effective safety policy under resource constraint, DOE (Design of experiment) was used. DOE uses classic designs, namely, fractional factorials and central composite designs. It used to make second order regression equation which serve as an objective function. That function was optimized under budget constraint and optimum value used for safety policy which shown greatest improvement in overall PSM. The outcome of this research indicates that personnel safety management model has the capability for acting as instruction tool to improve understanding of safety management and also as an aid to policy making

  6. Design an optimum safety policy for personnel safety management - A system dynamic approach

    Energy Technology Data Exchange (ETDEWEB)

    Balaji, P. [The Glocal University, Mirzapur Pole, Delhi- Yamuntori Highway, Saharanpur 2470001 (India)

    2014-10-06

    Personnel safety management (PSM) ensures that employee's work conditions are healthy and safe by various proactive and reactive approaches. Nowadays it is a complex phenomenon because of increasing dynamic nature of organisations which results in an increase of accidents. An important part of accident prevention is to understand the existing system properly and make safety strategies for that system. System dynamics modelling appears to be an appropriate methodology to explore and make strategy for PSM. Many system dynamics models of industrial systems have been built entirely for specific host firms. This thesis illustrates an alternative approach. The generic system dynamics model of Personnel safety management was developed and tested in a host firm. The model was undergone various structural, behavioural and policy tests. The utility and effectiveness of model was further explored through modelling a safety scenario. In order to create effective safety policy under resource constraint, DOE (Design of experiment) was used. DOE uses classic designs, namely, fractional factorials and central composite designs. It used to make second order regression equation which serve as an objective function. That function was optimized under budget constraint and optimum value used for safety policy which shown greatest improvement in overall PSM. The outcome of this research indicates that personnel safety management model has the capability for acting as instruction tool to improve understanding of safety management and also as an aid to policy making.

  7. Design an optimum safety policy for personnel safety management - A system dynamic approach

    Science.gov (United States)

    Balaji, P.

    2014-10-01

    Personnel safety management (PSM) ensures that employee's work conditions are healthy and safe by various proactive and reactive approaches. Nowadays it is a complex phenomenon because of increasing dynamic nature of organisations which results in an increase of accidents. An important part of accident prevention is to understand the existing system properly and make safety strategies for that system. System dynamics modelling appears to be an appropriate methodology to explore and make strategy for PSM. Many system dynamics models of industrial systems have been built entirely for specific host firms. This thesis illustrates an alternative approach. The generic system dynamics model of Personnel safety management was developed and tested in a host firm. The model was undergone various structural, behavioural and policy tests. The utility and effectiveness of model was further explored through modelling a safety scenario. In order to create effective safety policy under resource constraint, DOE (Design of experiment) was used. DOE uses classic designs, namely, fractional factorials and central composite designs. It used to make second order regression equation which serve as an objective function. That function was optimized under budget constraint and optimum value used for safety policy which shown greatest improvement in overall PSM. The outcome of this research indicates that personnel safety management model has the capability for acting as instruction tool to improve understanding of safety management and also as an aid to policy making.

  8. An innovative fuel design concept for improved Light Water Reactor performance and safety

    International Nuclear Information System (INIS)

    Tulenko, J.S.; Connell, R.G.

    1993-01-01

    The primary goal of this research is to develop a new fuel design which will have improved thermal/mechanical performance characteristics greatly superior to current thermal and mechanical design performance. The mechanical/thermal constraints define the lifetime of the fuel, the maximum power at which the fuel can be operated, the probability of fuel failure over core lifetime, and the integrity of a core during a transient excursion. The thermal/mechanical limits act to degrade fuel integrity when they are violated. The purpose of this project is to investigate a novel design for light water reactor fuel which will extend fuel performance limits and improve reactor safety even further than is currently achieved. This project is investigating liquid metal bonding of LWR fuel in order to radically decrease fuel centerline temperatures which has major performance and safety benefits. The project will verify the compatibility of the liquid metal bond with both the fuel pellets and cladding material, verify the performance enhancement features of the new design over the fuel lifetime, and verify the economic fabricability of the concept and will show how this concept will benefit the LWR nuclear industry

  9. LMFBR plant design features for sodium spill and fire protection

    International Nuclear Information System (INIS)

    Palm, R.E.

    1982-01-01

    Design features have been developed for an LMFBR plant to protect the concrete structures from potential liquid spills and fires and prevent sodium-concrete reactions. The inclusion of these features in the plant design reduces the severity of design basis accident conditions imposed on containment and other critical plant structures. Steel liners are provided in cells containing radioactive sodium systems, and catch pans are located in non-radioactive sodium system cells. The design requirements and descriptions of each of these protective features are presented. The loading conditions, analytical approach and numerical results are also included. Design of concrete cell structures that are subject to high temperature effects from sodium spills is discussed. The structural design considers the influence of high temperature on design properties of concrete and carbon steel materials based on results of a comprehensive test program. The development of these design features and high temperature design considerations for the Clinch River Breeder Reactor Plant (CRBRP) are presented in this paper

  10. Safety and licensing of nuclear heating plants

    International Nuclear Information System (INIS)

    Snell, V.G.; Hilborn, J.W.; Lynch, G.F.; McAuley, S.J.

    1989-09-01

    World attention continues to focus on nuclear district heating, a low-cost energy from a non-polluting fuel. It offers long-term security for countries currently dependent on fossil fuels, and can reduce the burden of fossil fuel transportation on railways and roads. Current initiatives encompass large, centralized heating plants and small plants supplying individual institutions. The former are variants of their power reactor cousins but with enhanced safety features. The latter face the safety and licensing challenges of urban siting and remotely monitored operation, through use of intrinsic safety features such as passive decay heat removal, low stored energy and limited reactivity speed and depth in the control systems. Small heating reactor designs are compared, and the features of the SLOWPOKE Energy System, in the forefront of these designs, are summarized. The challenge of public perception must be met by clearly presenting the characteristics of small heating reactors in terms of scale and transparent safety in design and operation, and by explaining the local benefits

  11. Safety Aspects of Sustainable Storage Dams and Earthquake Safety of Existing Dams

    Directory of Open Access Journals (Sweden)

    Martin Wieland

    2016-09-01

    Full Text Available The basic element in any sustainable dam project is safety, which includes the following safety elements: ① structural safety, ② dam safety monitoring, ③ operational safety and maintenance, and ④ emergency planning. Long-term safety primarily includes the analysis of all hazards affecting the project; that is, hazards from the natural environment, hazards from the man-made environment, and project-specific and site-specific hazards. The special features of the seismic safety of dams are discussed. Large dams were the first structures to be systematically designed against earthquakes, starting in the 1930s. However, the seismic safety of older dams is unknown, as most were designed using seismic design criteria and methods of dynamic analysis that are considered obsolete today. Therefore, we need to reevaluate the seismic safety of existing dams based on current state-of-the-art practices and rehabilitate deficient dams. For large dams, a site-specific seismic hazard analysis is usually recommended. Today, large dams and the safety-relevant elements used for controlling the reservoir after a strong earthquake must be able to withstand the ground motions of a safety evaluation earthquake. The ground motion parameters can be determined either by a probabilistic or a deterministic seismic hazard analysis. During strong earthquakes, inelastic deformations may occur in a dam; therefore, the seismic analysis has to be carried out in the time domain. Furthermore, earthquakes create multiple seismic hazards for dams such as ground shaking, fault movements, mass movements, and others. The ground motions needed by the dam engineer are not real earthquake ground motions but models of the ground motion, which allow the safe design of dams. It must also be kept in mind that dam safety evaluations must be carried out several times during the long life of large storage dams. These features are discussed in this paper.

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

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1989-01-01

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

  13. Key issues on safety design basis selection and safety assessment

    International Nuclear Information System (INIS)

    An, S.; Togo, Y.

    1976-01-01

    In current fast reactor design in Japan, four design accident conditions and four design seismic conditions are adopted as the design base classifications. These are classified by the considerations on both likelihood of occurrence and the severeness of the consequences. There are several major problem areas in safety design consideration such as core accident problems which include fuel sodium interaction, fuel failure propagation and residual decay heat removal, and decay heat removal systems problems which is more or less the problem of selection of appropriate system and of assurance of high reliability of the system. In view of licensing, two kinds of accidents are postulated in evaluating the adequacy of a reactor site. The one is the ''major accident'' which is the accident to give most severe radiation hazard to the public from technical point of view. The other is the ''hypothetical accident'', induced public accident of which is severer than that of major accident. While the concept of the former is rather unique to Japanese licensing, the latter is almost equivalent to design base hypothetical accident of the US practice. In this paper, design bases selections, key safety issues and some of the licensing considerations in Japan are described

  14. Development of safety analysis technology for integral reactor; evaluation on safety concerns of integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hee Chul; Kim, Woong Sik; Lee, J. H. [Korea Institute of Nuclear Safety, Taejeon (Korea)

    2002-03-01

    The Nuclear Desalination Plant (NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in this study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current light water reactor and advanced reactor designs, and user requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified and discussed. They include the use of proven technology for new safety features, systematic event classification and selection, strengthening containment function, and the safety impacts on desalination-related systems. The study presents the general safety requirements applicable to licensing of an integral reactor and suggests additional regulatory requirements, which need to be developed, based on the direction to resolution of the safety concerns. The efforts to identify and technically resolve the safety concerns in the design stage will provide the early confidence of SMART safety and the technical basis to evaluate the safety to designers and reviewers in the future. Suggestion on the development of additional regulatory requirements will contribute for the regulator to taking actions for licensing of an integral reactor. 66 refs., 5 figs., 24 tabs. (Author)

  15. The Westinghouse AP1000 plant design: a generation III+ reactor with unique proven passive safety technology

    International Nuclear Information System (INIS)

    Demetri, K. J.; Leipner, C. I.; Marshall, M. L.

    2015-09-01

    The AP1000 plant is an 1100-M We pressurized water reactor with passive safety features and extensive plant simplifications and standardization that simplify construction, operation, maintenance, safety, and cost. The AP1000 plant is based on proven pressurized water reactor (PWR) technology, with an emphasis on safety features that rely solely on natural forces. These passive safety features are combined with simple, active, defense-in-depth systems used during normal plant operations which also provide the first level of defense against more probable events. This paper focuses on specific safety and licensing topics: the AP1000 plant robustness to be prepared for extreme events that may lead to catastrophic loss of infrastructure, such as the Fukushima Dai-ichi event, and the AP1000 plant compliance with the safety objectives for new plants. The first deployment of the AP1000 plant formally began in July 2007 when Westinghouse Electric Company and its consortium partner, the Shaw Group, signed contracts for four AP1000 units on coastal sites of Sanmen and Haiyang, China. Both sites have the planned ability to accommodate at least six AP1000 units; construction is largely concurrent for all four units. Additionally, the United States (U.S.) Nuclear Regulatory Commission (NRC) issued combined licenses (COLs) to allow Southern Nuclear Operating Company (SNC) and South Carolina Electric and Gas Company (SCE and G) to construct and operate AP1000 plants. Within this paper, the various factors that contribute to an unparalleled level of design, construction, delivery, and licensing certainty for any new AP1000 plant projects are described. These include: 1) How the AP1000 plant design development and reviews undertaken in the United States, China and Europe increase licensing certainty. 2) How the AP1000 passive plant robustness against extreme events that result in large loss of infrastructure further contributes to the licensing certainty in a post

  16. The Westinghouse AP1000 plant design: a generation III+ reactor with unique proven passive safety technology

    Energy Technology Data Exchange (ETDEWEB)

    Demetri, K. J.; Leipner, C. I.; Marshall, M. L., E-mail: demetrkj@westinghouse.com [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2015-09-15

    The AP1000 plant is an 1100-M We pressurized water reactor with passive safety features and extensive plant simplifications and standardization that simplify construction, operation, maintenance, safety, and cost. The AP1000 plant is based on proven pressurized water reactor (PWR) technology, with an emphasis on safety features that rely solely on natural forces. These passive safety features are combined with simple, active, defense-in-depth systems used during normal plant operations which also provide the first level of defense against more probable events. This paper focuses on specific safety and licensing topics: the AP1000 plant robustness to be prepared for extreme events that may lead to catastrophic loss of infrastructure, such as the Fukushima Dai-ichi event, and the AP1000 plant compliance with the safety objectives for new plants. The first deployment of the AP1000 plant formally began in July 2007 when Westinghouse Electric Company and its consortium partner, the Shaw Group, signed contracts for four AP1000 units on coastal sites of Sanmen and Haiyang, China. Both sites have the planned ability to accommodate at least six AP1000 units; construction is largely concurrent for all four units. Additionally, the United States (U.S.) Nuclear Regulatory Commission (NRC) issued combined licenses (COLs) to allow Southern Nuclear Operating Company (SNC) and South Carolina Electric and Gas Company (SCE and G) to construct and operate AP1000 plants. Within this paper, the various factors that contribute to an unparalleled level of design, construction, delivery, and licensing certainty for any new AP1000 plant projects are described. These include: 1) How the AP1000 plant design development and reviews undertaken in the United States, China and Europe increase licensing certainty. 2) How the AP1000 passive plant robustness against extreme events that result in large loss of infrastructure further contributes to the licensing certainty in a post

  17. Generic radiation safety design for SSRL synchrotron radiation beamlines

    Energy Technology Data Exchange (ETDEWEB)

    Liu, James C. [Radiation Protection Department, Stanford Linear Accelerator Center (SLAC), MS 48, P.O. Box 20450, Stanford, CA 94309 (United States)]. E-mail: james@slac.stanford.edu; Fasso, Alberto [Radiation Protection Department, Stanford Linear Accelerator Center (SLAC), MS 48, P.O. Box 20450, Stanford, CA 94309 (United States); Khater, Hesham [Radiation Protection Department, Stanford Linear Accelerator Center (SLAC), MS 48, P.O. Box 20450, Stanford, CA 94309 (United States); Prinz, Alyssa [Radiation Protection Department, Stanford Linear Accelerator Center (SLAC), MS 48, P.O. Box 20450, Stanford, CA 94309 (United States); Rokni, Sayed [Radiation Protection Department, Stanford Linear Accelerator Center (SLAC), MS 48, P.O. Box 20450, Stanford, CA 94309 (United States)

    2006-12-15

    To allow for a conservative, simple, uniform, consistent, efficient radiation safety design for all SSRL beamlines, a generic approach has been developed, considering both synchrotron radiation (SR) and gas bremsstrahlung (GB) hazards. To develop the methodology and rules needed for generic beamline design, analytic models, the STAC8 code, and the FLUKA Monte Carlo code were used to pre-calculate sets of curves and tables that can be looked up for each beamline safety design. Conservative beam parameters and standard targets and geometries were used in the calculations. This paper presents the SPEAR3 beamline parameters that were considered in the design, the safety design considerations, and the main pre-calculated results that are needed for generic shielding design. In the end, the rules and practices for generic SSRL beamline design are summarized.

  18. Designation of facility usage categories for Hanford Site facilities

    International Nuclear Information System (INIS)

    Woodrich, D.D.; Ellingson, D.R.; Scott, M.A.; Schade, A.R.

    1991-10-01

    This report summarizes the Hanford Site methodology used to ensure facility compliance with the natural phenomena design criteria set forth in the US Department of Energy Orders and guidance. The current Hanford Site methodology for Usage Category designation is based on an engineered feature's safety function and on the feature's assigned Safety Class. At the Hanford Site, Safety Class assignments are deterministic in nature and are based on teh consequences of failure, without regard to the likelihood of occurrence. The report also proposes a risk-based approach to Usage Category designation, which is being considered for future application at the Hanford Site. To establish a proper Usage Category designation, the safety analysis and engineering design processes must be coupled. This union produces a common understanding of the safety function(s) to be accomplished by the design feature(s) and a sound basis for the assignment of Usage Categories to the appropriate systems, structures, and components. 4 refs., 9 figs., 1 tab

  19. Advanced nuclear power systems: Design, technology, safety and strategies for their deployment

    International Nuclear Information System (INIS)

    1994-01-01

    The objectives of the symposium were to provide high level decision makers with an overview of the discussion concerning the need for nuclear power and salient features of advanced nuclear power systems; a forum for discussing the design objectives and safety approaches for such systems and the views of regulatory bodies; a forum for identifying barriers to the deployment of these systems and for reviewing strategies to overcome these barriers; and a forum for reviewing options for international cooperation in the development and deployment of such systems. Refs, figs and tabs

  20. Safety design guides for seismic requirements for CANDU 9

    International Nuclear Information System (INIS)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; A. C. D. Wright

    1996-03-01

    This safety design guide for seismic requirements for CANDU 9 describes the seismic design philosophy, defines the applicable earthquakes and identifies the structures and systems requiring seismic qualification to ensure that the essential safety function can be adequately satisfied following earthquake. The detailed requirements for structures, systems and components which must be seismically qualified are specified in the Appendix. The change status of the regulatory requirements, code and standards should be traced and this safety design guide shall be updated accordingly. 1 fig., (Author) .new

  1. Technical basis for the ITER-FEAT outline design. Progress in resolving open design issues from the outline design report

    International Nuclear Information System (INIS)

    2000-01-01

    In this publication the technical basis for the ITER-FEAT outline design is presented. It comprises the Plant Design Specifications, the Safety Principles and Environmental Criteria, the Site Requirements and Site Design Assumptions. The outline of the key features of the ITER-FEAT design includes main physical parameters and assessment, design overview and preliminary safety assessment, cost and schedule

  2. Intermediate probabilistic safety assessment approach for safety critical digital systems

    International Nuclear Information System (INIS)

    Taeyong, Sung; Hyun Gook, Kang

    2001-01-01

    Even though the conventional probabilistic safety assessment methods are immature for applying to microprocessor-based digital systems, practical needs force to apply it. In the Korea, UCN 5 and 6 units are being constructed and Korean Next Generation Reactor is being designed using the digital instrumentation and control equipment for the safety related functions. Korean regulatory body requires probabilistic safety assessment. This paper analyzes the difficulties on the assessment of digital systems and suggests an intermediate framework for evaluating their safety using fault tree models. The framework deals with several important characteristics of digital systems including software modules and fault-tolerant features. We expect that the analysis result will provide valuable design feedback. (authors)

  3. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 2: Appendices

    International Nuclear Information System (INIS)

    1994-07-01

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR section 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE's application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design

  4. Westinghouse AP1000 advanced passive plant: design features and benefits

    International Nuclear Information System (INIS)

    Walls, S.J.; Cummins, W.E.

    2003-01-01

    The Westinghouse AP1000 Program is aimed at implementing the AP1000 plant to provide a further major improvement in plant economics while maintaining the passive safety advantages established by the AP600. An objective is to retain to the maximum extent possible the plant design of the AP600 so as to retain the licensing basis, cost estimate, construction schedule, modularization scheme, and the detailed design from the AP600 program. Westinghouse and the US Nuclear Regulatory Commission staff have embarked on a program to complete Design Certification for the AP1000 by 2004. A pre-certification review phase was completed in March 2002 and was successful in establishing the applicability of the AP600 test program and AP600 safety analysis codes to the AP1000 Design Certification. On March 28, 2002, Westinghouse submitted to US NRC the AP1000 Design Control Document and Probabilistic Risk Assessment, thereby initiating the formal design certification review process. The results presented in these documents verify the safety performance of the API 000 and conformance with US NRC licensing requirements. Plans are being developed for implementation of a series of AP1000 plants in the US. Key factors in this planning are the economics of AP1000, and the associated business model for licensing, constructing and operating these new plants. Similarly plans are being developed to get the AP1000 design reviewed for use in the UK. Part of this planning has been to examine the AP1000 design relative to anticipated UK safety and licensing issues. (author)

  5. Safety Design Criteria (SDC) for Gen-IV Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Nakai, Ryodai

    2013-01-01

    SDC Development Background & Objectives: • Safety Design Criteria (SDC) Development for Gen-IV SFR: – Proposed at the GIF Policy Group (PG) meeting in October 2010 –SDC “harmonization” is increasingly important for: • Realization of enhanced safety designs meeting to Gen-IV safety goals and safety approach common to SFR systems; • Preparation for the forthcoming licensing in the near future; • Because Gen-IV SFR are progressing into conceptual design stage. • The SDC is the Reference criteria: – Of the designs of safety-related Structures, Systems & Components that are specific to the SFR system; – For clarifying the requisites systematically & comprehensively; – When the technology developers apply the basic safety approach and use the codes & standards for conceptual design of the Gen-IV SFR system

  6. Improving the design of higher-capacity railway tank cars for hazardous materials transport: Optimizing the trade-off between weight and safety

    International Nuclear Information System (INIS)

    Barkan, Christopher P.L.

    2008-01-01

    As with many aspects of modern industrial society, decision-makers face trade-offs in considering hazardous materials transportation equipment and practices. Tank cars used for transport of hazardous materials can be made more resistant to damage in accidents through use of a thicker steel tank and other protective features. However, the additional weight of these features reduces the car's capacity and thus its efficiency as a transportation vehicle. In this paper the problem of tank car safety versus weight is developed as a multi-attribute decision problem. North American railroads recently developed specifications for higher capacity tank cars for transportation of hazardous materials including enhanced safety design features. A group of tank car safety design features or 'risk reduction options' (RROs) were analyzed with regard to their effect on the conditional probability of release in an accident, and their incremental effect on tank car weight. All possible combinations of these RROs were then analyzed in terms of the reduced release probability per unit of weight increase and the Pareto optimal set of options identified. This set included the combinations of RROs that provided the greatest improvement in safety with the least amount of additional weight for any desired level of tank car weight increase. The analysis was conducted for both non-insulated and insulated tank cars and used two objective functions, minimization of conditional probability of release, and minimization of expected quantity lost, given that a car was derailed in an accident. Sensitivity analyses of the effect of tank car size and use of different objective functions were conducted and the optimality results were found to be robust. The results of this analysis were used by the Association of American Railroads Tank Car Committee to develop new specifications for higher capacity non-insulated and insulated, non-pressure tank cars resulting in an estimated 32% and 24% respective

  7. Engineering features of the INTOR conceptual design

    International Nuclear Information System (INIS)

    Shannon, T.E.

    1981-01-01

    The INTOR engineering design has been strongly influenced by considerations for assembly and maintenance. A maintenance philosophy was established at the outset of the conceptual design to insure that the tokamak configuration would be developed to accommodate maintenance requirements. The main features of the INTOR design are summarized in this paper with primary emphasis on the impact of maintenance considerations

  8. Safety upgrading of the PAKS Nuclear Plant

    International Nuclear Information System (INIS)

    Vamos, G.; Vigassy, J.

    1993-01-01

    In the last several years the net electricity from the Paks NPP represents almost half of the Hungarian total. The 4 units of Paks belong to the latest generation of the VVER-440 units, the small-sized Russian designed PWRs. Reviewing the main design features of them, the safety merits and safety concerns are summarized. Due to the conservative design and the extensive operating experience the safety merits appear to be more significant than generally believed. The VVER-440 type has two models, the 230 and 213, which have a large number of distinctive safety features. These are highlighted in the section comparisons. A quality assurance program was initiated in Paks very early. A long-term safety upgrading program was also initiated, originating from vendor recommendations, regulatory decisions, in-house operating experience and safety concerns, and independent reviews. The main areas and some examples of the measures are described. This program, like all other activities related to nuclear safety, has been under regulatory control. The specific features of the Hungarian regulatory system are described. For advanced, general and new evaluation of the safety of the units in Paks in accordance with the internationally recommended criteria of the 90's, the project AGNES has been launched with international participation. The scope of this project is summarized. International efforts as the IAEA Regional Project on safety assessment of VVER-440/213 and VVER-440/230 units are underway. Since safety is not only a question of design, but it can be significantly influenced by operations and maintenance practices, the Paks NPP has invited LAEA's OSART and ASSET missions, WANO's Pilot Peer Review

  9. Space nuclear reactor safety

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  10. Design of reactor containment systems for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2008-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems

  11. Design of reactor containment systems for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems

  12. Definition and Means of Maintaining the Criticality Prevention Design Features Portion of the PFP Safety Envelope

    International Nuclear Information System (INIS)

    RAMBLE, A.L.

    2000-01-01

    The purpose of this document is to record the technical evaluation of the Operational Safety Requirements described in the Plutonium Finishing Plant Final (PFP) Operational Safety Requirements, WHC-SD-CP-OSR-010. Rev. 0-N , Section 3.1.1, ''Criticality Prevention System.'' This document, with its appendices, provides the following: (1) The results of a review of Criticality Safety Analysis Reports (CSAR), later called Criticality Safety Evaluation Reports (CSER), and Criticality Prevention Specifications (CPS) to determine which equipment or components analyzed in the CSER or CPS are considered as one of the two unlikely, independent, and concurrent changes before a criticality accident is possible. (2) Evaluations of equipment or components to determine the safety boundary for the system (Section 4). (3) A list of essential drawings that show the safety system or component (Appendix A). (4) A list of the safety envelope (SE) equipment (Appendix B). (5) Functional requirements for the individual safety envelope equipment (Sections 3 and 4). (6) A list of the operational and surveillance procedures necessary to maintain the system equipment within the safety envelope (Section 5)

  13. Design for safety: theoretical framework of the safety aspect of BIM system to determine the safety index

    Directory of Open Access Journals (Sweden)

    Ai Lin Evelyn Teo

    2016-12-01

    Full Text Available Despite the safety improvement drive that has been implemented in the construction industry in Singapore for many years, the industry continues to report the highest number of workplace fatalities, compared to other industries. The purpose of this paper is to discuss the theoretical framework of the safety aspect of a proposed BIM System to determine a Safety Index. An online questionnaire survey was conducted to ascertain the current workplace safety and health situation in the construction industry and explore how BIM can be used to improve safety performance in the industry. A safety hazard library was developed based on the main contributors to fatal accidents in the construction industry, determined from the formal records and existing literature, and a series of discussions with representatives from the Workplace Safety and Health Institute (WSH Institute in Singapore. The results from the survey suggested that the majority of the firms have implemented the necessary policies, programmes and procedures on Workplace Safety and Health (WSH practices. However, BIM is still not widely applied or explored beyond the mandatory requirement that building plans should be submitted to the authorities for approval in BIM format. This paper presents a discussion of the safety aspect of the Intelligent Productivity and Safety System (IPASS developed in the study. IPASS is an intelligent system incorporating the buildable design concept, theory on the detection, prevention and control of hazards, and the Construction Safety Audit Scoring System (ConSASS. The system is based on the premise that safety should be considered at the design stage, and BIM can be an effective tool to facilitate the efforts to enhance safety performance. IPASS allows users to analyse and monitor key aspects of the safety performance of the project before the project starts and as the project progresses.

  14. Designation of facility usage categories for Hanford Site facilities

    International Nuclear Information System (INIS)

    Wodrich, D.; Ellingson, D.; Scott, M.; Schade, A.

    1991-01-01

    This report summarizes the Hanford Site methodology used to ensure facility compliance with the natural phenomena design criteria set forth in the US Department of Energy orders and guidance. In particular, the Hanford Site approach to designating a suitable facility open-quotes Usage Category,close quotes is presented. The current Hanford Site methodology for Usage Category designation is based on an engineered feature's safety function and on the feature's assigned Safety Class. At the Hanford Site, Safety Class assignments are deterministic in nature and are based on the consequences of failure, without regard to the likelihood of occurrence. The report also proposes a risk-based approach to Usage Category designation, which is being considered for future application at the Hanford Site. To establish a proper Usage Category designation, the safety analysis and engineering design processes must be coupled. This union produces a common understanding of the safety function(s) to be accomplished by the design feature(s) and a sound basis for the assignment of Usage Categories to the appropriate systems, structures, and components

  15. Feature-based component model for design of embedded systems

    Science.gov (United States)

    Zha, Xuan Fang; Sriram, Ram D.

    2004-11-01

    An embedded system is a hybrid of hardware and software, which combines software's flexibility and hardware real-time performance. Embedded systems can be considered as assemblies of hardware and software components. An Open Embedded System Model (OESM) is currently being developed at NIST to provide a standard representation and exchange protocol for embedded systems and system-level design, simulation, and testing information. This paper proposes an approach to representing an embedded system feature-based model in OESM, i.e., Open Embedded System Feature Model (OESFM), addressing models of embedded system artifacts, embedded system components, embedded system features, and embedded system configuration/assembly. The approach provides an object-oriented UML (Unified Modeling Language) representation for the embedded system feature model and defines an extension to the NIST Core Product Model. The model provides a feature-based component framework allowing the designer to develop a virtual embedded system prototype through assembling virtual components. The framework not only provides a formal precise model of the embedded system prototype but also offers the possibility of designing variation of prototypes whose members are derived by changing certain virtual components with different features. A case study example is discussed to illustrate the embedded system model.

  16. European Workshop Industrical Computer Science Systems approach to design for safety

    Science.gov (United States)

    Zalewski, Janusz

    1992-01-01

    This paper presents guidelines on designing systems for safety, developed by the Technical Committee 7 on Reliability and Safety of the European Workshop on Industrial Computer Systems. The focus is on complementing the traditional development process by adding the following four steps: (1) overall safety analysis; (2) analysis of the functional specifications; (3) designing for safety; (4) validation of design. Quantitative assessment of safety is possible by means of a modular questionnaire covering various aspects of the major stages of system development.

  17. Nuclear reactor conceptual design: methodology for cost-effective internalisation of nuclear safety

    International Nuclear Information System (INIS)

    Gimenez, M.; Grinblat, P.; Schlamp, M.

    2002-01-01

    A novel and promising methodology to perform nuclear reactor design is presented in this work. It achieves to balance efficiently safety and economics at the conceptual engineering stage. The key to this integral approach is to take into account safety aspects in a design optimisation process where the design variables are balanced in order to obtain a better figure of merit related with reactor economic performance. Design parameter effects on characteristic or critical safety variables, chosen from reactor behaviour during accidents and from its probabilistic safety assessment -safety performance indicators-, are synthesised on Safety Design Maps. These maps allow one to compare these indicators with limit values, which are determined by design criteria or regulations, and to transfer these restrictions to the design parameters. In this way, reactor dynamic response and other safety aspects are integrated in a global optimisation process, by means of additional rules to the neutronic, thermal-hydraulic and mechanical calculations. This methodology turns out to be promising to balance and optimise reactor and safety system design in an early engineering stage, in order to internalise cost-efficiently safety issues. It also allows one to evaluate the incremental costs of implementing higher safety levels. Furthermore, through this methodology, a simplified design can be obtained, compared to the resultant complexity when these concepts are introduced in a later engineering stage. (author)

  18. Richards Barrier LA Reference Design Feature Evaluation

    International Nuclear Information System (INIS)

    N.E. Kramer

    1999-01-01

    The Richards Barrier is one of the design features of the repository to be considered for the License Application (LA), Richards was a soil scientist who first described the diversion of moisture between two materials with different hydrologic properties. In this report, a Richards Barrier is a special type of backfill with a fine-grained material (such as sand) overlaying a coarse-grained material (such as gravel). Water that enters an emplacement drift will first encounter the fine-grained material and be transported around the coarse-grained material covering the waste package, thus protecting the waste package from contact with most of the groundwater. The objective of this report is to discuss the benefits and liabilities to the repository by the inclusion of a Richards Barrier type backfill in emplacement drifts. The Richards Barrier can act as a barrier to water flow, can reduce the waste package material dissolution rate, limit mobilization of the radionuclides, and can provide structural protection for the waste package. The scope of this report is to: (1) Analyze the behavior of barrier materials following the intrusion of groundwater for influxes of 1 to 300 mm per year. The report will demonstrate diversion of groundwater intrusions into the barrier over an extended time period when seismic activity and consolidation may cause the potential for liquefaction and settlement of the Richards Barrier. (2) Review the thermal effects of the Richards Barrier on material behavior. (3) Analyze the effect of rockfall on the performance of the Richards Barrier and the depth of the barrier required to protect waste packages under the barrier. (4) Review radiological and heating conditions on placement of multiple layers of the barrier. Subsurface Nuclear Safety personnel will perform calculations to determine the radiation reduction-time relationship and shielding capacity of the barrier. (5) Evaluate the effects of ventilation on cooling of emplacement drifts and

  19. Passive safety and the advanced liquid metal reactors

    International Nuclear Information System (INIS)

    Hill, D.J.; Pedersen, D.R.; Marchaterre, J.F.

    1988-01-01

    Advanced Liquid Metal Reactors being developed today in the USA are designed to make maximum use of passive safety features. Much of the LMR safety work at Argonne National Laboratory is concerned with demonstrating, both theoretically and experimentally, the effectiveness of the passive safety features. The characteristics that contribute to passive safety are discussed, with particular emphasis on decay heat removal systems, together with examples of Argonne's theoretical and experimental programs in this area

  20. Code on the safety of nuclear research reactors: Design

    International Nuclear Information System (INIS)

    1992-01-01

    The main objective of this publication is to provide a safety basis for the design of a research reactor and for the assessment of the design. Another objective is to cover certain aspects related to regulatory supervision, siting and quality assurance, as far as these are related to activities for the design of a research reactor. These objectives are expressed in terms of requirements and recommendations for the design of research reactors. Emphasis is placed on the safety requirements that shall be met rather than on ways in which they can be met. The requirements and recommendations may form the foundation necessary for a Member State to develop specific regulations and safety criteria for its research reactor programme.

  1. Study on the KALIMER safety approach

    International Nuclear Information System (INIS)

    Kim, Eui Kwang; Han, Do Hee; Kim, Young Cheol.

    1997-01-01

    This study describes KALIMER's safety approach, how to establish the safety criteria and temperature limit, how to define safety evaluation events, and some safety research and development needs items. It is recommended that the KALIMER's approach to safety use seven levels of safety design and a defense-in-depth design approach with particular emphasis on inherent passive features. In order to establish as set DBEs for KALIMER safety evaluation, the procedure is explained how to define safety evaluation events. Final selection is to be determined later with the final establishment of design concepts. On the basis of preliminary studies and evaluation of the plant safety related areas, the KALIMER and PRISM have following three main difference that may require special research and development for KALIMER. (author). 7 refs., 6 tabs., 6 figs

  2. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Spanish Edition); Seguridad de las centrales nucleares: Diseno. Requisitos de seguridad especificos

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-04-15

    This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  3. Demonstration of inherent safety features of HTGRs using the HTTR

    International Nuclear Information System (INIS)

    Tachibana, Yukio; Nakagawa, Shigeaki; Nakazawa, Toshio; Iyoku, Tatsuo

    2004-01-01

    Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are conducted for the purpose of demonstrating inherent safety features of High Temperature Gas-cooled Reactors (HTGRs) quantitatively as well as providing the core and plant transient data for validation of HTGR analysis codes for safety evaluation. The safety demonstration test are divided to the first phase and second phase tests. In the first phase tests, simulation tests of anticipated operational occurrences and anticipated transients without scram (ATWS) are conducted. The second phase tests will simulate accidents such as a depressurization accident (loss of coolant accident). The first phase test simulating reactivity insertion events and coolant flow reduction events stared in FY 2002. Post-test analyses have been conducted to reproduced the test results by using the core and plant dynamics analysis code, ACCORD and Monte Carlo code, MVP. The analysis results agreed fairly well with the test results of a control rod withdrawal test simulating reactivity insertion, and gas circulators trip test simulating coolant flow reduction, at power levels of 50% and 30% of the rated power, respectively. It is shown that improvement of the ACCORD code by taking into consideration vertical and horizontal temperature distribution gives better analysis results in the control rod withdrawal test. The fist phase safety demonstration tests will continue until FY 2005, and the second phase tests are planned to be started in FY 2006. (author)

  4. Incorporation of Safety into Design Process : A Systems Engineering Perspective

    NARCIS (Netherlands)

    Rajabalinejad, M.

    2018-01-01

    This paper suggests integrating the best safety practices with the design process. This integration enriches the exploration experience for designers and adds extra values and competitor advantages for customers. The paper introduces the safety cube for combining common blocks for design, hazard

  5. Modeling crash injury severity by road feature to improve safety.

    Science.gov (United States)

    Penmetsa, Praveena; Pulugurtha, Srinivas S

    2018-01-02

    The objective of this research is 2-fold: to (a) model and identify critical road features (or locations) based on crash injury severity and compare it with crash frequency and (b) model and identify drivers who are more likely to contribute to crashes by road feature. Crash data from 2011 to 2013 were obtained from the Highway Safety Information System (HSIS) for the state of North Carolina. Twenty-three different road features were considered, analyzed, and compared with each other as well as no road feature. A multinomial logit (MNL) model was developed and odds ratios were estimated to investigate the effect of road features on crash injury severity. Among the many road features, underpass, end or beginning of a divided highway, and on-ramp terminal on crossroad are the top 3 critical road features. Intersection crashes are frequent but are not highly likely to result in severe injuries compared to critical road features. Roundabouts are least likely to result in both severe and moderate injuries. Female drivers are more likely to be involved in crashes at intersections (4-way and T) compared to male drivers. Adult drivers are more likely to be involved in crashes at underpasses. Older drivers are 1.6 times more likely to be involved in a crash at the end or beginning of a divided highway. The findings from this research help to identify critical road features that need to be given priority. As an example, additional advanced warning signs and providing enlarged or highly retroreflective signs that grab the attention of older drivers may help in making locations such as end or beginning of a divided highway much safer. Educating drivers about the necessary skill sets required at critical road features in addition to engineering solutions may further help them adopt safe driving behaviors on the road.

  6. Safety design of SNR-300

    International Nuclear Information System (INIS)

    Traube, K.

    1976-01-01

    The joint German-Belgian-Dutch loop-type 300 MW(e) LMFBR prototype is being constructed at Kalkar on the lower Rhine in Germany. Among the many arguments put forward in defense of SNR-300, that of acquiring licensing exprience has proven to be of major importance to the international breeder scene. The severity of the licensing procedures and of the safety standards imposed are unique in several respects, including timing: generally growing scepticism towards nuclear power increased severity of the licensing practice; organizational features: the procedure and criteria developed for commercial light water reactors have been applied without exemptions. This relates to the commercial-type contract under which SNR-300 is being built for private utilities by a private company; and German nuclear safety standards, known worldwide to be most stringent. The following three important areas are discussed in which SNR-300 decidedly deviates from its forerunners: protection against the hypothetical core disruptive accident (HCDA), protection against external events, and provisions for in-service inspection

  7. Technical features of ABWR safety systems

    International Nuclear Information System (INIS)

    Sugisaki, Toshihiko; Tominaga, Kenji; Horiuchi, Tetsuo

    1986-01-01

    The engineering safety facilities of ABWRs have been disigned so as to have many excellent characteristics such as safety, reliability and economy, reflecting the merit of adopting new technology such as internal pumps and new control rod driving mechanism, and coupled with the safety peculiar to BWRs. In this paper, about ECCS, containment vessels and others which compose the engineering safety facilities of ABWRs, the characteristics related to the safety owing to the adoption of internal pumps and others, and the evaluation of the performance at the time of various accidents are discussed. As the results of safety evaluation, it was clarified that due to the safety peculiar to ABWRs and the characteristics of the safety facilities, the large increases of safety, reliability and economy have been planned in the ABWRs, and for example, core flooding can be maintained even at the time of a hypothetical loss of coolant accident. BWRs have the simple system constitution, good self controllability, large natural circulation ability, simple operation control method and excellent ability of confining heat and radioactivity. BWRs have three safety functions to stop reactors, to remove heat from reactors, and to confine radioactive substances. These functions of ABWRs were evaluated, and very high safety was confirmed. (Kako, I.)

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

    International Nuclear Information System (INIS)

    G. L. Sharp; R. T. McCracken

    2004-01-01

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

  9. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 1: Main report

    International Nuclear Information System (INIS)

    1994-07-01

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR section 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE's application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design

  10. Final safety evaluation report related to the certification of the advanced boiling water reactor design. Volume 1: Main report

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-01

    This safety evaluation report (SER) documents the technical review of the US Advanced Boiling Water Reactor (ABWR) standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the ABWR design was initially submitted by the General Electric Company, now GE Nuclear Energy (GE), in accordance with the procedures of Appendix O of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50). Later GE requested that its application be considered as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. The ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated power of 3,926 megawatts thermal (MWt) and a design power of 4,005 MWt. To the extent feasible and appropriate, the staff relied on earlier reviews for those ABWR design features that are substantially the same as those previously considered. Unique features of the ABWR design include internal recirculation pumps, fine-motion control rod drives, microprocessor-based digital logic and control systems, and digital safety systems. On the basis of its evaluation and independent analyses, the NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the US ABWR standard design.

  11. Generic Safety Issue (GSI) 171 -- Engineered Safety Feature (ESF) failure from a loop subsequent to LOCA: Assessment of plant vulnerability and CDF contributions

    International Nuclear Information System (INIS)

    Martinez-Guridi, G.; Samanta, P.; Chu, L.; Yang, J.

    1998-01-01

    Generic Safety Issue 171 (GSI-171), Engineered Safety Feature (ESF) from a Loss Of Offsite Power (LOOP) subsequent to a Loss Of Coolant Accident (LOCA), deals with an accident sequence in which a LOCA is followed by a LOOP. This issue was later broadened to include a LOOP followed by a LOCA. Plants are designed to handle a simultaneous LOCA and LOOP. In this paper, the authors address the unique issues that are involved i LOCA with delayed LOOP (LOCA/LOOP) and LOOP with delayed LOCA (LOOP/LOCA) accident sequences. LOCA/LOOP accidents are analyzed further by developing event-tree/fault-tree models to quantify their contributions to core-damage frequency (CDF) in a pressurized water reactor and a boiling water reactor (PWR and a BWR). Engineering evaluation and judgments are used during quantification to estimate the unique conditions that arise in a LOCA/LOOP accident. The results show that the CDF contribution of such an accident can be a dominant contributor to plant risk, although BWRs are less vulnerable than PWRs

  12. Safety design guides for grouping and separation for CANDU 9

    International Nuclear Information System (INIS)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; A. C. D. Wright

    1996-03-01

    This safety design guide for grouping and separation describes the philosophy of physical and functional separation for systems, structures and components in CANDU 9 plants and provides the requirements for the implementation of the philosophy in the detailed plant design. The separation of the safety systems is to ensure that common cause events and functional interconnections between systems do not impair the capability to perform the required safety functions for accident conditions. The separation requirements are also applied to the design by grouping the plant systems into two basic groups. Group 1 includes the power production systems and Group 2 includes the safety related systems required for the mitigation of serious process failure. The Group 2 is further separated into subgroups to ensure that events that could cause failure of a special safety system in one subgroup can be mitigated by the other subgroup. The change status for the regulatory requirements, code and standards should be traced and this safety design guide shall be updated accordingly. 2 tabs., 6 figs. (Author) .new

  13. Safety design guides for grouping and separation for CANDU 9

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young [Korea Atomic Energy Research Institute, Daeduk (Korea, Republic of); Wright, A C.D. [Atomic Energy of Canada Ltd., Toronto (Canada)

    1996-03-01

    This safety design guide for grouping and separation describes the philosophy of physical and functional separation for systems, structures and components in CANDU 9 plants and provides the requirements for the implementation of the philosophy in the detailed plant design. The separation of the safety systems is to ensure that common cause events and functional interconnections between systems do not impair the capability to perform the required safety functions for accident conditions. The separation requirements are also applied to the design by grouping the plant systems into two basic groups. Group 1 includes the power production systems and Group 2 includes the safety related systems required for the mitigation of serious process failure. The Group 2 is further separated into subgroups to ensure that events that could cause failure of a special safety system in one subgroup can be mitigated by the other subgroup. The change status for the regulatory requirements, code and standards should be traced and this safety design guide shall be updated accordingly. 2 tabs., 6 figs. (Author) .new.

  14. Reliability Improved Design for a Safety System Channel

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Eung Se; Kim, Yun Goo [KHNP, Daejeon (Korea, Republic of)

    2016-05-15

    Nowadays, these systems are implemented with a same platform type, such as a qualified programmable logic controller (PLC). The platform intensively uses digital communication with fiber-optic links to reduce cabling costs and to achieve effective signal isolation. These communication interface and redundancies within a channel increase the complexness of an overall system design. This paper proposes a simpler channel architecture design to reduce the complexity and to enhance overall channel reliability. Simplified safety channel configuration is proposed and the failure probabilities are compared with baseline safety channel configuration using an estimated generic value. The simplified channel configuration achieves 40 percent failure reduction compare to baseline safety channel configuration. If this configuration can be implemented within a processor module, overall safety channel reliability is increase and costs of fabrication and maintenance will be greatly reduced.

  15. Reliability Improved Design for a Safety System Channel

    International Nuclear Information System (INIS)

    Oh, Eung Se; Kim, Yun Goo

    2016-01-01

    Nowadays, these systems are implemented with a same platform type, such as a qualified programmable logic controller (PLC). The platform intensively uses digital communication with fiber-optic links to reduce cabling costs and to achieve effective signal isolation. These communication interface and redundancies within a channel increase the complexness of an overall system design. This paper proposes a simpler channel architecture design to reduce the complexity and to enhance overall channel reliability. Simplified safety channel configuration is proposed and the failure probabilities are compared with baseline safety channel configuration using an estimated generic value. The simplified channel configuration achieves 40 percent failure reduction compare to baseline safety channel configuration. If this configuration can be implemented within a processor module, overall safety channel reliability is increase and costs of fabrication and maintenance will be greatly reduced

  16. Safety principles and design management of Chashma Nuclear Power Plant

    International Nuclear Information System (INIS)

    Geng Qirui; Cheng Pingdong

    1997-01-01

    The basic safety consideration and detailed design principles in the design of Chashma Nuclear Power Plant is elaborated. The management within the frame setting up by 'safety culture' and 'quality culture'

  17. Design of integrated passive safety system (IPSS) for ultimate passive safety of nuclear power plants

    International Nuclear Information System (INIS)

    Chang, Soon Heung; Kim, Sang Ho; Choi, Jae Young

    2013-01-01

    Highlights: • We newly propose the design concept of integrated passive safety system (IPSS). • It has five safety functions for decay heat removal and severe accident mitigation. • Simulations for IPSS show that core melt does not occur in accidents with SBO. • IPSS can achieve the passive in-vessel retention and ex-vessel cooling strategy. • The applicability of IPSS is high due to the installation outside the containment. -- Abstract: The design concept of integrated passive safety system (IPSS) which can perform various passive safety functions is proposed in this paper. It has the various functions of passive decay heat removal system, passive safety injection system, passive containment cooling system, passive in-vessel retention and cavity flooding system, and filtered venting system with containment pressure control. The objectives of this paper are to propose the conceptual design of an IPSS and to estimate the design characters of the IPSS with accident simulations using MARS code. Some functions of the IPSS are newly proposed and the other functions are reviewed with the integration of the functions. Consequently, all of the functions are modified and integrated for simplicity of the design in preparation for beyond design based accidents (BDBAs) focused on a station black out (SBO). The simulation results with the IPSS show that the decay heat can be sufficiently removed in accidents that occur with a SBO. Also, the molten core can be retained in a vessel via the passive in-vessel retention strategy of the IPSS. The actual application potential of the IPSS is high, as numerous strong design characters are evaluated. The installation of the IPSS into the original design of a nuclear power plant requires minimal design change using the current penetrations of the containment. The functions are integrated in one or two large tanks outside the containment. Furthermore, the operation time of the IPSS can be increased by refilling coolant from the

  18. Safety considerations in the design of the fusion engineering device

    International Nuclear Information System (INIS)

    Barrett, R.J.

    1983-01-01

    Safety considerations play a significant role in the design of a near-term Fusion Engineering Device (FED). For the safety of the general public and the plant workers, the radiation environment caused by the reacting plasma and the potential release of tritium fuel are the dominant considerations. The U.S. Department of Energy (DOE) regulations and guidelines for radiation protection have been reviewed and are being applied to the device design. Direct radiation protection is provided by the device shield and the reactor building walls. Radiation from the activated device components and the tritium fuel is to be controlled with shielding, contamination control, and ventilation. The potential release of tritium from the plant has influenced the selection of reactor building and plant designs and specifications. The safety of the plant workers is affected primarily by the radiation from the activated device components and from plasma chamber debris. The highly activated device components make it necessary to design many of the maintenance activities in the reactor building for totally remote operation. The hot cell facility has evolved as a totally remote maintenance facility due to the high radiation levels of the device components. Safety considerations have had substantial impacts on the design of FED. Several examples of safety-related design impacts are discussed in the paper. Feasible solutions have been identified for all outstanding safety-related items, and additional optimization of these solutions is anticipated in future design studies

  19. Design Information from the PSA for Digital Safety-Critical Systems

    International Nuclear Information System (INIS)

    Kang, Hyun Gook; Jang, Seung Cheol

    2005-01-01

    Many safety-critical applications such as nuclear field application usually adopt a similar design strategy for digital safety-critical systems. Their differences from the normal design for the non-safety-critical applications could be summarized as: multiple-redundancy, highly reliable components, strengthened monitoring mechanism, verified software, and automated test procedure. These items are focusing on maintaining the capability to perform the given safety function when it is requested. For the past several decades, probabilistic safety assessment (PSA) techniques are used in the nuclear industry to assess the relative effects of contributing events on plant risk and system reliability. They provide a unifying means of assessing physical faults, recovery processes, contributing effects, human actions, and other events that have a high degree of uncertainty. The applications of PSA provide not only the analysis results of already installed system but also the useful information for the system under design. The information could be derived from the PSA experience of the various safety-critical systems. Thanks to the design flexibility, the digital system is one of the most suitable candidates for risk-informed design (RID). In this article, we will describe the feedbacks for system design and try to develop a procedure for RID. Even though the procedure is not sophisticated enough now, it could be the start point of the further investigation for developing more complete and practical methodology

  20. Analysis and design on airport safety information management system

    Directory of Open Access Journals (Sweden)

    Yan Lin

    2017-01-01

    Full Text Available Airport safety information management system is the foundation of implementing safety operation, risk control, safety performance monitor, and safety management decision for the airport. The paper puts forward the architecture of airport safety information management system based on B/S model, focuses on safety information processing flow, designs the functional modules and proposes the supporting conditions for system operation. The system construction is helpful to perfecting the long effect mechanism driven by safety information, continually increasing airport safety management level and control proficiency.

  1. Safety design guides for environmental qualification for CANDU 9

    International Nuclear Information System (INIS)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; A. C. D. Wright

    1996-03-01

    This safety design guide describes the safety philosophy and requirements for the environmental qualification of safety related systems and components for CANDU 9. The environmental qualification program identifies the equipments to be qualified and conditions to be used for qualification and provides comprehensive set of documentation to ensure that the qualification is complete and can be maintained for the life of the plant. A summary of the system, components and structures requiring environmental qualification is provided in the table for the guidance of the system design, and this table will be subject to change or confirmation by the environmental qualification program. Also, plant ares subject to harsh environment is provided in the figure. The change status of the regulatory requirements, code and standards should be traced and this safety design guide shall be updated accordingly. 1 tab., 5 figs. (Author) .new

  2. Safety characteristics of the US advanced liquid metal reactor core

    International Nuclear Information System (INIS)

    Magee, P.M.; Dubberley, A.E.; Gyorey, G.L.; Lipps, A.J.; Wu, T.

    1991-01-01

    The U.S. Advanced Liquid Metal Reactor (ALMR) design employs innovative, passive features to provide an unprecedented level of public safety and the ability to demonstrate this safety to the public. The key features employed in the core design to produce the desired passive safety characteristics are: a small core with a tight restraint system, the use of metallic U-Pu-Zr fuel, control rod withdrawal limiters, and gas expansion modules. In addition, the reactor vessel and closure are designed to have the capability to withstand, with large margins, the maximum possible core disruptive accident without breach and radiological release. (author)

  3. Transient safety performance of the PRISM innovative liquid metal reactor

    International Nuclear Information System (INIS)

    Magee, P.M.; Dubberley, A.E.; Rhow, S.K.; Wu, T.

    1988-01-01

    The PRISM sodium-cooled reactor concept utilizes passive safety characteristics and modularity to increase performance margins, improve licensability, reduce owner's risk and reduce costs. The relatively small size of each reactor module (471 MWt) facilitates the use of passive self-shutdown and shutdown heat removal features, which permit design simplification and reduction of safety-related systems. Key to the transient performance is the inherent negative reactivity feedback characteristics of the core design resulting from the use of metal (U-Pu-Zr) swing, and very low control rod runout worth. Selected beyond design basis events relying only on these core design features are analyzed and the design margins summarized to demonstrate the advancement in reactor safety achieved with the PRISM design concept

  4. Review of design criteria and safety analysis of safety class electric building for fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1998-02-01

    Steady state fuel test loop will be equipped in HANARO to obtain the development and betterment of advanced fuel and materials through the irradiation tests. HANARO fuel test loop was designed for CANDU and PWR fuel testing. Safety related system of Fuel Test Loop such as emergency cooling water system, component cooling water system, safety ventilation system, high energy line break mitigation system and remote control room was required 1E class electric supply to meet the safety operation in accordance with related code. Therefore, FTL electric building was designed to construction and install the related equipment based on seismic category I. The objective of this study is to review the design criteria and analysis the safety function of safety class electric building for fuel test loop, and this results will become guidance for the irradiation testing in future. (author). 10 refs., 6 tabs., 30 figs.

  5. Simplified safety and containment systems for the iris reactor

    International Nuclear Information System (INIS)

    Conway, L.E.; Lombardi, C.; Ricotti, M.; Oriani, L.

    2001-01-01

    The IRIS (International Reactor Innovative and Secure) is a 100 - 300 MW modular type pressurized water reactor supported by the U.S. DOE NERI Program. IRIS features a long-life core to provide proliferation resistance and to reduce the volume of spent fuel, as well as reduce maintenance requirements. IRIS utilizes an integral reactor vessel that contains all major primary system components. This integral reactor vessel makes it possible to reduce containment size; making the IRIS more cost competitive. IRIS is being designed to enhance reactor safety, and therefore a key aspect of the IRIS program is the development of the safety and containment systems. These systems are being designed to maximize containment integrity, prevent core uncover following postulated accidents, minimize the probability and consequences of severe accidents, and provide a significant simplification over current safety system designs. The design of the IRIS containment and safety systems has been identified and preliminary analyses have been completed. The IRIS safety concept employs some unique features that minimize the consequences of postulated design basis events. This paper will provide a description of the containment design and safety systems, and will summarize the analysis results. (author)

  6. Safety requirements in the design of research reactors: A Canadian perspective

    International Nuclear Information System (INIS)

    Lee, A.G.; Langman, V.J.

    2000-01-01

    In Canada, the formal development of safety requirements for the design of research reactors in general began under an inter-organizational Small Reactor Criteria Committee. This committee developed safety and licensing criteria for use by several small reactor projects in their licensing discussions with the Atomic Energy Control Board. The small reactor projects or facilities represented included the MAPLE-X10 reactor, the proposed SES-10 heating reactor and its prototype, the SDR reactor at the Whiteshell Laboratories, the Korea Multipurpose Research Reactor (a.k.a., HANARO) in Korea, the SCORE project, and the McMaster University Nuclear Reactor. The top level set of criteria which form a safety philosophy and serve as a framework for more detailed developments was presented at an IAEA Conference in 1989. AECL continued this work to develop safety principles and design criteria for new small reactors. The first major application of this work has been to the design, safety analysis and licensing of the MAPLE 1 and 2 reactors for the MDS Nordion Medical Isotope Reactor Project. This paper provides an overview of the safety principles and design criteria. Examples of an implementation of these safety principles and design criteria are drawn from the work to design the MAPLE 1 and 2 reactors. (author)

  7. Evaluation of the Ventilation and Air Cleaning System Design Concepts for Safety Requirements during Fire Conditions in Nuclear Applications

    International Nuclear Information System (INIS)

    Rashad, S.; El-Fawal, M.; Kandil, M.

    2013-01-01

    The ventilation and air cleaning system in the nuclear or radiological installations is one of the essential nuclear safety concerns. It is responsible for confining the radioactive materials involved behind suitable barriers during normal and abnormal conditions. It must be designed to prevent the release of harmful products (radioactive gases, or airborne radioactive materials) from the system or facility, impacting the public or workers, and doing environmental damage. There are two important safety functions common to all ventilation and air cleaning system in nuclear facilities. They are: a) the requirements to maintain the pressure of the ventilated volume below that of surrounding, relatively non-active areas, in order to inhibit the spread of contamination during normal and abnormal conditions, and b) the need to treat the ventilated gas so as to minimize the release of any radioactive or toxic materials. Keeping the two important safety functions is achieved by applying the fire protection for the ventilation system to achieve safety and adequate protection in nuclear applications facilities during fire and accidental criticality conditions.The main purpose of this research is to assist ventilation engineers and experts in nuclear installations for safe operation and maintaining ventilation and air cleaning system during fire accident in nuclear facilities. The research focuses on fire prevention and protection of the ventilation systems in nuclear facilities. High-Efficiency particulate air (HEPA) filters are extremely susceptible to damage when exposed to the effects of fire, smoke, and water; it is the intent of this research to provide the designer with the experience gained over the years from hard lessons learned in protecting HEPA filters from fire. It describes briefly and evaluates the design safety features, constituents and working conditions of ventilation and air cleaning system in nuclear and radioactive industry.This paper provides and

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

    International Nuclear Information System (INIS)

    Cahalana, J. E.; Hahn, D.

    2007-01-01

    This report summarizes the results of a three-year collaboration between Argonne National Laboratory (ANL) and the Korea Atomic Energy Research Institute (KAERI) to identify and quantify the performance of innovative design features in metallic-fueled, sodium-cooled fast reactor designs. The objective of the work was to establish the reliability and safety margin enhancements provided by design innovations offering significant potential for construction, maintenance, and operating cost reductions. The project goal was accomplished with a combination of advanced model development (Task 1), analysis of innovative design and safety features (Tasks 2 and 3), and planning of key safety experiments (Task 4). Task 1--Computational Methods for Analysis of Passive Safety Design Features: An advanced three-dimensional subassembly thermal-hydraulic model was developed jointly and implemented in ANL and KAERI computer codes. The objective of the model development effort was to provide a high-accuracy capability to predict fuel, cladding, coolant, and structural temperatures in reactor fuel subassemblies, and thereby reduce the uncertainties associated with lower fidelity models previously used for safety and design analysis. The project included model formulation, implementation, and verification by application to available reactor tests performed at EBR-II. Task 2--Comparative Analysis and Evaluation of Innovative Design Features: Integrated safety assessments of innovative liquid metal reactor designs were performed to quantify the performance of inherent safety features. The objective of the analysis effort was to identify the potential safety margin enhancements possible in a sodium-cooled, metal-fueled reactor design by use of passive safety mechanisms to mitigate low-probability accident consequences. The project included baseline analyses using state-of-the-art computational models and advanced analyses using the new model developed in Task 1. Task 3--Safety

  9. Safety design study of fast breeder reactors in Japan

    International Nuclear Information System (INIS)

    Miura, M.; Inagaki, T.

    1992-01-01

    This paper reports on two fast breeder reactor (FBR) concepts, the tank type and the loop type, that have been studied as possible reactor designs to be used for a demonstration FBR (DFBR). The basic principle fo the DFBR design is to ensure plant safety through a defense-in-depth methodology. Improvements in the seismic and thermal stress designs have been attempted for both reactor concepts. The system design study strives to maximize the reliability of the safety-related systems and to rationalize commercialization of the plant

  10. Preliminary safety analysis of unscrammed events for KLFR

    International Nuclear Information System (INIS)

    Kim, S.J.; Ha, G.S.

    2005-01-01

    The report presents the design features of KLFR; Safety Analysis Code; steady-state calculation results and analysis results of unscrammed events. The calculations of the steady-state and unscrammed events have been performed for the conceptual design of KLFR using SSC-K code. UTOP event results in no fuel damage and no centre-line melting. The inherent safety features are demonstrated through the analysis of ULOHS event. Although the analysis of ULOF has much uncertainties in the pump design, the analysis results show the inherent safety characteristics. 6% flow of rated flow of natural circulation is formed in the case of ULOF. In the metallic fuel rod, the cladding temperature is somewhat high due to the low heat transfer coefficient of lead. ULOHS event should be considered in design of RVACS for long-term cooling

  11. Safety design and evaluation policy for future FBRs in Japan

    International Nuclear Information System (INIS)

    Aizawa, Kiyoto

    1991-01-01

    The safety policy for fast breeder reactors (FBRs) has gradually matured in accordance with the development of FBRs. The safety assessment of the Japanese prototype FBR, Monju during the licensing process accelerated the maturity and the integration of knowledge and databases. Results are expected to be reflected in the establishment of the safety design and evaluation policy for FBRs. Although the methodologies and safety policies developed for LWRs are applicable in principle to future FBRs, it is neither rational nor realistic to treat safety only with these policies. It is recommended that one should develop the methodologies and safety policies starting from understanding of the inherent safety characteristics of FBR's through safety research, plant operating experience and design work. In the last few years, some technical committees were organized in Japan and have discussed key safety issues which are specific to FBRs in order to provide preparatory reports and to establish safety standards and guidelines for future commercial FBRs. (author)

  12. Safety design integrated in the Building Delivery System

    DEFF Research Database (Denmark)

    Jørgensen, Kirsten

    2012-01-01

    phases of the building delivery system by using the principle of the lean construction modelling. The method for the research was to go through the lean construction building delivery system step by step and create a normative description of what to do, when to do and how to do to fully integration...... of safety in each process. The group of participants who created the description had a high experience in a combination of research, safety and health in general and especial in construction and knowledge of the lean construction processes both from the clients perspective as well as from the designers...... and the consultants. The result is a concept and guideline including control schemes for how to integrate safety design in the lean construction building delivery system including what to do and when. The concept has been tested in an educational context and found useful by the designers. The practical value...

  13. Innovative small and medium sized reactors: Design features, safety approaches and R and D trends. Final report of a technical meeting

    International Nuclear Information System (INIS)

    2005-05-01

    In order to beat the economy of scale small and medium sized reactors (SMRs) have to incorporate specific design features that result into simplification of the overall plant design, modularization and mass production. Several approaches are being under development and consideration, including the increased use of passive features for reactivity control and reactor shut down, decay heat removal and core cooling, and reliance on the increased margin to fuel failure achieved through the use of advanced high-temperature fuel forms and structural materials. Some SMRs also offer the possibility of very long core lifetimes with burnable absorbers or high conversion ratio in the core. These reactors incorporate increased proliferation resistance and may offer a very attractive solution for the implementation of adequate safeguards in a scenario of global deployment of nuclear power. About 50 concepts and designs of the innovative SMRs are under development in more than 15 IAEA Member States representing both industrialized and developing countries. SMRs are under development for all principle reactor lines, i.e., water cooled, liquid metal cooled, gas cooled, and molten salt cooled reactors, as well as for some non-conventional combinations thereof. Upon a diversity of the conceptual and design approaches to SMRs, it may be useful to identify the so-called enabling technologies that are common to certain reactor types or lines. An enabling technology is the technology that needs to be developed and demonstrated to make a certain reactor concept viable. When a certain technology is common to several SMR concepts or designs, it could benefit from being developed on a common or shared basis. The identification of common enabling technologies could speed up the development and deployment of many SMRs by merging the efforts of their designers through an increased international cooperation. This publication has been prepared through the collaboration of all participants of this

  14. Proposal for a technology-neutral safety approach for new reactor designs

    International Nuclear Information System (INIS)

    2007-09-01

    Many states are considering an expansion of their nuclear power generation programmes. Many of the technologies and concepts are new and innovative. The current design and licensing rules are applicable to mostly large water reactors and there are no accepted rules in place for design, safety assessment and licensing for new innovative nuclear power plants. This TECDOC proposes a (new) safety approach and a methodology to generate technology-neutral (i.e. independent of reactor technology) safety requirements and a 'safe design' for advanced and innovative reactors. The experience gained in decades of design and licensing, combined with the development of risk-based concepts, has provided insights that will form the basis for new safety rules and requirements. Many lessons learned acknowledge the importance of such concepts as safety goals and defence in depth and the benefits of integrating risk insights early in an iterative design process. A new safety approach will incorporate many of the new developments in these concepts. For example, the probabilistic elements of defence in depth will help define the cumulative provisions to compensate for uncertainty and incompleteness of our knowledge of accident initiation and progression. This TECDOC also identifies areas of work, which will require further definition, research and development and guidance on application. This publication is to be used as a guide to developing a new technology-neutral safety approach, and as a guide in the application of methodologies to define the safety requirements for an innovative reactor designs. The method proposes an integration of deterministic and probabilistic considerations with established principles and concepts such as safety goals and defence in depth. The TECDOC recommends that the structure of the new technology-neutral main pillars for the design and licensing of innovative nuclear reactors be developed following a top-down approach to reflect a newer risk-informed and

  15. Development of advanced design features for KNGR reactor vessel and internals

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Kyun; Ru, Bong; Lee, Jae Han; Lee, Hyung Yeon; Kim, Jong Bum; Ku, Kyung Heoy; Lee, Ki Young; Lee, Jun; Kim, Young In [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-12-01

    Developments of KNGR design require to enhance the design to implement the design requirements, such as plant life time from 40 years to 60 years, safety, performance and structure and components design. The designs used for existing nuclear power plants should be modified or improved to meet the requirements in KNGR design. The purpose of the task is to develop the Advanced Design Features (ADF) related to mechanical and structural design for KNGR reactor vessel and reactor internals. The structural integrity for the System 80+ reactor vessel, of which design life is 60 years, was reviewed. EPRI-URD, CESSAR-DC, and the present design status and characteristics of System 80+ reactor vessel were comparatively studied and the improvement of reactor vessel surveillance program was investigated. The performance and aseismic characteristics of the CE-type CEDM, which will be used in System 80+, are investigated. The driving cycles of CEDM are evaluated for the load follow operation(LFO), of which Mode K is being developed by KAERI. The position of the USNRC, EPRI, ABB-CE, and industries on the elimination of OBE are reviewed, and especially ABB-CE System 80+ FSER is reviewed in detail. For the pre-stage of the verification of the OBE elimination from the design, the review of the seismic responses, i.e.. shear forces and moments, of YGN 3/4 RI was performed and the ratio of OBE response to SSE response was analysed. The screening criteria were reviewed to evaluate the integrity against pressurized thermal shock (PTS) for RV belt-line of System 80+. The evaluation methods for fracture integrity when screening criteria are not met were reviewed. The structural characteristics of IRWST spargers of System 80+ were investigated and the effect of hydrodynamic loads on NSSS was reviewed. 18 figs., 9 tabs., 40 refs. (Author) .new.

  16. Development of advanced design features for KNGR reactor vessel and internals

    International Nuclear Information System (INIS)

    Park, Jong Kyun; Ru, Bong; Lee, Jae Han; Lee, Hyung Yeon; Kim, Jong Bum; Ku, Kyung Heoy; Lee, Ki Young; Lee, Jun; Kim, Young In

    1995-12-01

    Developments of KNGR design require to enhance the design to implement the design requirements, such as plant life time from 40 years to 60 years, safety, performance and structure and components design. The designs used for existing nuclear power plants should be modified or improved to meet the requirements in KNGR design. The purpose of the task is to develop the Advanced Design Features (ADF) related to mechanical and structural design for KNGR reactor vessel and reactor internals. The structural integrity for the System 80+ reactor vessel, of which design life is 60 years, was reviewed. EPRI-URD, CESSAR-DC, and the present design status and characteristics of System 80+ reactor vessel were comparatively studied and the improvement of reactor vessel surveillance program was investigated. The performance and aseismic characteristics of the CE-type CEDM, which will be used in System 80+, are investigated. The driving cycles of CEDM are evaluated for the load follow operation(LFO), of which Mode K is being developed by KAERI. The position of the USNRC, EPRI, ABB-CE, and industries on the elimination of OBE are reviewed, and especially ABB-CE System 80+ FSER is reviewed in detail. For the pre-stage of the verification of the OBE elimination from the design, the review of the seismic responses, i.e.. shear forces and moments, of YGN 3/4 RI was performed and the ratio of OBE response to SSE response was analysed. The screening criteria were reviewed to evaluate the integrity against pressurized thermal shock (PTS) for RV belt-line of System 80+. The evaluation methods for fracture integrity when screening criteria are not met were reviewed. The structural characteristics of IRWST spargers of System 80+ were investigated and the effect of hydrodynamic loads on NSSS was reviewed. 18 figs., 9 tabs., 40 refs. (Author) .new

  17. Design and safety aspects of nuclear district heating reactors

    International Nuclear Information System (INIS)

    Brogli, R.; Mathews, D.; Pelloni, S.

    1989-01-01

    Extensive studies on the rationale, the potential and the technology of nuclear district heating have been performed in Switzerland. Beside economics the safety aspects were of primary importance. Due to the high costs to transport heat the heating reactor tend to be small and therefore, minimally staffed and located close to population centers. Stringed safety rules are therefore applying. Gas cooled reactors are well suited as district heating reactors since they have due to their characteristics several inherent features, significant safety margins and a remarkable radioactivity retention potential. Some ways to mitigate the effects of water ingress and graphite corrosion are under investigation. (author). 5 refs, 3 figs

  18. Safety design philosophy of the ABWR for the next generation LWRs

    International Nuclear Information System (INIS)

    Sato, Takashi; Akinaga, Makoto; Kojima, Yoshihiro

    2009-01-01

    The paper presents safety design philosophy of the advanced boiling water reactor (ABWR) to be reflected in developing the next generation light water reactors (LWRs). The basic policy of the ABWR safety design was to improve safety and reduce cost simultaneously by reflecting lessons learned of precursors, incidents and accidents that were beyond the design basis such as the Three Mile Island Unit 2 (TMI 2) accident. The ABWR is a fully active safety plant. The ABWR enhanced redundancy and diversity of active safety systems using probabilistic safety assessment (PSA) insights. It adopted a complete three division active emergency core cooling system (ECCS) and attained a very low core damage frequency (CDF) value of less than 10 -7 /ry for internal events. Only very small residual risks, if any, rather exist in external events such as an extremely large earthquake beyond the design basis. This is because external events can constitute a common cause that disables all the redundant active safety systems. Therefore, it is useless to add one more ECCS train and make a four division active ECCS for external events. Nowadays, however, fully passive safety LWRs are already established. Incorporating some of these passive safety systems we can also establish the next generation LWRs that are truly strong against external events. We can establish a plant that can survive a giant earthquake at least three days without AC power source, SA proof safety design that enables no containment failure and no evacuation to eliminate the residual risks. The same basic policy as the ABWR to improve safety and reduce cost simultaneously is again effective for the next generation LWRs. (author)

  19. U.S. ALMR safety approach and licensing status

    International Nuclear Information System (INIS)

    Hardy, R.W.; Gyorey, G.L.

    1991-01-01

    The Advanced Liquid Metal Cooled Reactor in the United States is based on the PRISM concept originated by General Electric. This concept features a compact modular system suitable for factory fabrication, and a high degree of passive and natural safety characteristics. The safety approach emphasizes accident prevention, backed up by accident mitigation as required. First-round safety evaluations by the U.S. regulators have found that the design provides passive, natural and other desirable features enhancing the safety of the power plant. Licensing review continuing. (author)

  20. Safety analyses for NHR-200

    Energy Technology Data Exchange (ETDEWEB)

    Jincai, Li; Zuying, Gao; Baocheng, Xu; Junxiao, He [Institute of Nuclear Energy and Technology, Tsingua Univ., Beijing (China)

    1997-09-01

    The NHR-200 is a commercial 200-MW District Heating Reactor developed in China. It is designed on the basis of design, construction and four-year operating experience of the 5MW Experimental Heating Reactor (NHR-5). It has special safety features which are briefly described in this paper. Accident classification and safety criteria are also explained. Some typical and serious accidents are studied theoretically, and their results are detailed in this paper. They demonstrate the excellent safety characteristics of HR-200. (author). 4 refs, 9 figs, 1 tab.

  1. Programmable Electronic Safety Systems

    International Nuclear Information System (INIS)

    Parry, R.

    1993-05-01

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

  2. Safety design analyses of Korea Advanced Liquid Metal Reactor

    International Nuclear Information System (INIS)

    Suk, S.D.; Park, C.K.

    2000-01-01

    The national long-term R and D program updated in 1997 requires Korea Atomic Energy Research Institute (KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor (KALIMER), along with supporting R and D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 MWe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self consistent design meeting a set of the major safety design requirements for accident prevention. Some of current emphasis include those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve supporting R and D programs of substance. This paper summarizes some of the results of engineering and design analyses performed for the safety of KALIMER. (author)

  3. The ACR: Advanced design features for a short construction schedule

    International Nuclear Information System (INIS)

    Elgohary, M.; Fairclough, N.

    2003-01-01

    Building on the successful CANDU construction at Qinshan, the ACR-700 is designed with constructability considerations as a major requirement during all project phases from the concept design stage to the detail design stage. A project schedule of 48 months has been developed for the nth ACR unit with a 36 months construction period from First Concrete to Fuel Load. This paper describes some of the advanced design features implemented in the reactor building design in order to achieve this short construction period. These features include large volume concrete pours, prefabricated rebar, composite structures, prefabricated permanent formwork and significant modularization and prefabrication

  4. Cold Vacuum Drying Safety Class Instrumentation and Control System Design Description SYS 93-2

    International Nuclear Information System (INIS)

    WHITEHURST, R.

    1999-01-01

    This document describes the Cold Vacuum Drying Facility (CVDF) Safety Class Instrumentation and Control system (SCIC). The SCIC provides safety functions and features to protect the environment, off-site and on-site personnel and equipment. The function of the SCIC is to provide automatic trip features, valve interlocks, alarms, indication and control for the cold vacuum drying process

  5. Safety of research reactors (Design and Operation)

    International Nuclear Information System (INIS)

    Dirar, H. M.

    2012-06-01

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

  6. European passive plant program preliminary safety analyses to support system design

    International Nuclear Information System (INIS)

    Saiu, Gianfranco; Barucca, Luciana; King, K.J.

    1999-01-01

    In 1994, a group of European Utilities, together with Westinghouse and its Industrial Partner GENESI (an Italian consortium including ANSALDO and FIAT), initiated a program designated EPP (European Passive Plant) to evaluate Westinghouse Passive Nuclear Plant Technology for application in Europe. In the Phase 1 of the European Passive Plant Program which was completed in 1996, a 1000 MWe passive plant reference design (EP1000) was established which conforms to the European Utility Requirements (EUR) and is expected to meet the European Safety Authorities requirements. Phase 2 of the program was initiated in 1997 with the objective of developing the Nuclear Island design details and performing supporting analyses to start development of Safety Case Report (SCR) for submittal to European Licensing Authorities. The first part of Phase 2, 'Design Definition' phase (Phase 2A) was completed at the end of 1998, the main efforts being design definition of key systems and structures, development of the Nuclear Island layout, and performing preliminary safety analyses to support design efforts. Incorporation of the EUR has been a key design requirement for the EP1000 form the beginning of the program. Detailed design solutions to meet the EUR have been defined and the safety approach has also been developed based on the EUR guidelines. The present paper describes the EP1000 approach to safety analysis and, in particular, to the Design Extension Conditions that, according to the EUR, represent the preferred method for giving consideration to the Complex Sequences and Severe Accidents at the design stage without including them in the design bases conditions. Preliminary results of some DEC analyses and an overview of the probabilistic safety assessment (PSA) are also presented. (author)

  7. Specific features of medicines safety and pharmacovigilance in Africa

    Science.gov (United States)

    Pal, Shanthi N.; Olsson, Sten; Dodoo, Alexander; Bencheikh, Rachida Soulayami

    2012-01-01

    The thalidomide tragedy in the late 1950s and early 1960s served as a wakeup call and raised questions about the safety of medicinal products. The developed countries rose to the challenge putting in place systems to ensure the safety of medicines. However, this was not the case for low-resource settings because of prevailing factors inherent in them. This paper reviews some of these features and the current status of pharmacovigilance in Africa. The health systems in most of the 54 countries of Africa are essentially weak, lacking in basic infrastructure, personnel, equipment and facilities. The recent mass deployment of medicines to address diseases of public health significance in Africa poses additional challenges to the health system with notable safety concerns. Other safety issues of note include substandard and counterfeit medicines, medication errors and quality of medicinal products. The first national pharmacovigilance centres established in Africa with membership of the World Health Organization (WHO) international drug monitoring programme were in Morocco and South Africa in 1992. Of the 104 full member countries in the programme, there are now 24 African countries with a further nine countries as associate members. The pharmacovigilance systems operational in African countries are based essentially on spontaneous reporting facilitated by the introduction of the new tool Vigiflow. The individual case safety reports committed to the WHO global database (Vigibase) attest to the growth of pharmacovigilance in Africa with the number of reports rising from 2695 in 2000 to over 25,000 in 2010. There is need to engage the various identified challenges of the weak pharmacovigilance systems in the African setting and to focus efforts on how to provide resources, infrastructure and expertise. Raising the level of awareness among healthcare providers, developing training curricula for healthcare professionals, provisions for paediatric and geriatric

  8. Fusion reactor passive safety and ignitor risk-based regulation

    International Nuclear Information System (INIS)

    Zucchetti, M.

    1995-01-01

    Passive design features are more reliable than operator action of successful operation of active safety systems. Passive safety has usually been adopted for fission. The achievement of an inventory-based passive safety is difficult if the fusion reactor uses neutronic reactions. Ignitor is a high-magnetic field tokamak designed to study the physics of ignited plasmas. The safety goal for Ignitor is classification as a mobility-based passively safe machine

  9. Identification of features of electronic prescribing systems to support quality and safety in primary care using a modified Delphi process.

    Science.gov (United States)

    Sweidan, Michelle; Williamson, Margaret; Reeve, James F; Harvey, Ken; O'Neill, Jennifer A; Schattner, Peter; Snowdon, Teri

    2010-04-15

    Electronic prescribing is increasingly being used in primary care and in hospitals. Studies on the effects of e-prescribing systems have found evidence for both benefit and harm. The aim of this study was to identify features of e-prescribing software systems that support patient safety and quality of care and that are useful to the clinician and the patient, with a focus on improving the quality use of medicines. Software features were identified by a literature review, key informants and an expert group. A modified Delphi process was used with a 12-member multidisciplinary expert group to reach consensus on the expected impact of the features in four domains: patient safety, quality of care, usefulness to the clinician and usefulness to the patient. The setting was electronic prescribing in general practice in Australia. A list of 114 software features was developed. Most of the features relate to the recording and use of patient data, the medication selection process, prescribing decision support, monitoring drug therapy and clinical reports. The expert group rated 78 of the features (68%) as likely to have a high positive impact in at least one domain, 36 features (32%) as medium impact, and none as low or negative impact. Twenty seven features were rated as high positive impact across 3 or 4 domains including patient safety and quality of care. Ten features were considered "aspirational" because of a lack of agreed standards and/or suitable knowledge bases. This study defines features of e-prescribing software systems that are expected to support safety and quality, especially in relation to prescribing and use of medicines in general practice. The features could be used to develop software standards, and could be adapted if necessary for use in other settings and countries.

  10. Safety criteria for design of nuclear power plants

    International Nuclear Information System (INIS)

    1997-01-01

    In Finland the general safety requirements for nuclear power plants are presented in the Council of State Decision (395/91). In this guide, safety principles which supplement the Council of State Decision and which are to be used in the design of nuclear power plants are defined

  11. AP600 - an ALWR conceptual design

    International Nuclear Information System (INIS)

    Bruce, R.A.; Vijuk, R.P.

    1988-01-01

    The Electric Power Research Institute is spearheading an effort to develop utility requirements for the Advanced Light Water Reactor (ALWR) plants which will become the next generation nuclear power plants for the U.S. This EPRI ALWR Program involves utilities, the U.S. Department of Energy, the U.S. Nuclear Regulatory Commission, and various industry suppliers. The ALWR Program is aimed at ALWR plants which incorporate step improvements in safety, reliability, operability and power generation costs. As part of the ALWR efforts, a Westinghouse team is conducting conceptual design development of a PWR plant design called the AP600, reflecting advanced passive safety features and the chosen 600 MWe plant output. The AP600 conceptual design provides significant improvements while employing proven component technology. This paper describes the basic reactor and primary coolant system features, the passive safety system features, and plant arrangement/construction features of AP600

  12. Safety design concept and analysis for the upgrading JRR-3

    International Nuclear Information System (INIS)

    Onishi, N.; Isshiki, M.; Takahashi, H.; Takayanagi, M.

    1990-01-01

    The Research Reactor No.3 (JRR-3) is under reconstruction for upgrading. This paper describes the safety design concepts of the architectural and engineering design, anticipated operational transients and accident conditions which are the postulated initiating events for the safety evaluation, and the safety criteria of the upgraded JRR-3. The safety criteria are defined taking into account those of Light Water Reactors and the characteristics of the research reactor. Using the example of the safety analysis, this paper describes analytical results of a reactivity insertion by removal of in-core irradiation samples, a pipeline break at the primary coolant loop and flow blockage to a coolant channel, which are the severest postulated initiating events of the JRR-3

  13. Diffusive Barrier and Getter Under Waste Packages VA Reference Design Feature Evaluations

    International Nuclear Information System (INIS)

    MacNeil, K.

    1999-01-01

    This technical document evaluates those aspects of the diffusive barrier and getter features which have the potential for enhancing the performance of the Viability Assessment Reference Design and are also directly related to the key attributes for the repository safety strategy of that design. The effects of advection, hydrodynamic dispersion, and diffusion on the radionuclide migration rates through the diffusive barrier were determined through the application of the one-dimensional, advection/dispersion/diffusion equation. The results showed that because advective flow described by the advection-dispersion equation dominates, the diffusive barrier feature alone would not be effective in retarding migration of radiocuclides. However, if the diffusive barrier were combined with one or more features that reduced the potential for advection, then transport of radionuclides would be dominated by diffusion and their migration from the EBS would be impeded. Apatite was chosen as the getter material used for this report. Two getter configurations were developed, Case 1 and Case 2. As in the evaluation of the diffusive barrier, the effects of advection, hydrodynamic dispersion, and diffusion on the migration of radionuclides through the getter are evaluated. However, in addition to these mechanisms, the one-dimensional advection/dispersion/diffusion model is modified to include the effect of sorption on radionuclide migration rates through the sorptive medium (getter). As a result of sorption, the longitudinal dispersion coefficient, and the average linear velocity are effectively reduced by the retardation factor. The retardation factor is a function of the getter material's dry bulk density, sorption coefficient and moisture content. The results of the evaluation showed that a significant delay in breakthrough through the getter can be achieved if the thickness of the getter barrier is increased

  14. Conceptual safety design analysis of Korea advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Suk, S. D.; Park, C. K.

    1999-01-01

    The national long-term R and D program, updated in 1977, requires Korea Atomic Energy Research Institute (KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor (KALIMER), along with supporting R and D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 Mwe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self-consistent design meeting a set of major safety design requirements for accident prevention. Some of the current emphasis includes those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve extensive supporting R and D programs. This paper summarizes some of the results of conceptual engineering and design analyses performed for the safety of KALIMER in the area of inherent safety, passive decay heat removal, sodium water reaction, and seismic isolation. (author)

  15. Regulatory issues resolved through design certification on the System 80+trademark standard plant design

    International Nuclear Information System (INIS)

    Ritterbusch, S.E.; Brinkman, C.B.

    1996-01-01

    The US Nuclear Regulatory Commission (NRC) has completed its review of the System 80+trademark Standard Plant Design, approving advanced design features and closing severe accident licensing issues. Final Design Approval was granted in July 1994. The NRC review was extensive, requiring written responses to over 4,950 questions and formal printing of over 50,000 Safety Analysis Report pages. New safety issues never before addressed in a regulatory atmosphere had to be resolved with detailed analysis and evaluation of design features. the System 80+ review demonstrated that regulatory issues can be firmly resolved only through presentation of a detailed design and completion of a comprehensive regulatory review

  16. Safety principles and design criteria for nuclear power stations

    International Nuclear Information System (INIS)

    Gazit, M.

    1982-01-01

    The criteria and safety principles for the design of nuclear power stations are presented from the viewpoint of a nuclear engineer. The design, construction and operation of nuclear power stations should be carried out according to these criteria and safety principles to ensure, to a reasonable degree, that the likelihood of release of radioactivity as a result of component failure or human error should be minimized. (author)

  17. Safety Design Criteria of Indian Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  18. Final safety evaluation report related to the certification of the System 80+ design (Docket No. 52-002). Volume 1, Chapters 1--14

    International Nuclear Information System (INIS)

    1994-08-01

    This final safety evaluation report (FSER) documents the technical review of the System 80+ standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the System 80+ design was submitted by Combustion Engineering, Inc., now Asea Brown Boveri-Combustion Engineering (ABB-CE) as an application for design approval and subsequent design certification pursuant to 10 CFR section 52.45. System 80+ is a pressurized water reactor with a rated power of 3914 megawatts thermal (MWt) and a design power of 3992 MWt at which accidents are analyzed. Many features of the System 80+ are similar to those of Abb-CE's System 80 design from which it evolved. Unique features of the System 80+ design included: a large spherical, steel containment; an in-containment refueling water storage tank; a reactor cavity flooding system, hydrogen ignitors, and a safety depressurization system for severe accident mitigation; a combustion gas turbine for an alternate ac source; and an advanced digitally based control room. On the basis of its evaluation and independent analyses, the NRC staff concludes that ABB-CE's application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the System 80+ standard design. This document, Volume 1, contains Chapters 1 through 14 of this report

  19. Status of the EU test blanket systems safety studies

    International Nuclear Information System (INIS)

    Panayotov, Dobromir; Poitevin, Yves; Ricapito, Italo; Zmitko, Milan

    2015-01-01

    Highlights: • TBS safety demonstration files. • Safety functions and related design features – detailed TBS components classifications. • Nuclear analyses, radiation shielding and protection. • TBS radiological waste management strategy and categorization. • Selection and definition of reference accidents scenarios and accidents analyses. - Abstract: The European joint undertaking for ITER and the development of fusion energy (‘Fusion for Energy’ – F4E) provides the European contributions to the ITER international fusion energy research project. Among others it includes also the development, design, technological demonstration and implementation of the European test blanket systems (TBS) in ITER. Currently two EU TBS designs are in the phase of conceptual design – helium-cooled lithium-lead (HCLL) and helium-cooled pebble-bed (HCPB). Safety demonstration is an important part of the work devoted to the achievement of the next key project milestone the conceptual design review. The paper reveals the details of the work on EU TBS safety performed in the last couple of years: update of the TBS safety demonstration files; safety functions and related design features; detailed TBS components classifications; nuclear analyses, radiation shielding and protection; TBS radiological waste management strategy and categorization; selection and definition of reference accidents scenarios, and accidents analyses. Finally the authors share the information on on-going and planned future EU TBS safety activities.

  20. Status of the EU test blanket systems safety studies

    Energy Technology Data Exchange (ETDEWEB)

    Panayotov, Dobromir, E-mail: dobromir.panayotov@f4e.europa.eu; Poitevin, Yves; Ricapito, Italo; Zmitko, Milan

    2015-10-15

    Highlights: • TBS safety demonstration files. • Safety functions and related design features – detailed TBS components classifications. • Nuclear analyses, radiation shielding and protection. • TBS radiological waste management strategy and categorization. • Selection and definition of reference accidents scenarios and accidents analyses. - Abstract: The European joint undertaking for ITER and the development of fusion energy (‘Fusion for Energy’ – F4E) provides the European contributions to the ITER international fusion energy research project. Among others it includes also the development, design, technological demonstration and implementation of the European test blanket systems (TBS) in ITER. Currently two EU TBS designs are in the phase of conceptual design – helium-cooled lithium-lead (HCLL) and helium-cooled pebble-bed (HCPB). Safety demonstration is an important part of the work devoted to the achievement of the next key project milestone the conceptual design review. The paper reveals the details of the work on EU TBS safety performed in the last couple of years: update of the TBS safety demonstration files; safety functions and related design features; detailed TBS components classifications; nuclear analyses, radiation shielding and protection; TBS radiological waste management strategy and categorization; selection and definition of reference accidents scenarios, and accidents analyses. Finally the authors share the information on on-going and planned future EU TBS safety activities.

  1. Design of plant safety model in plant enterprise engineering environment

    International Nuclear Information System (INIS)

    Gabbar, Hossam A.; Suzuki, Kazuhiko; Shimada, Yukiyasu

    2001-01-01

    Plant enterprise engineering environment (PEEE) is an approach aiming to manage the plant through its lifecycle. In such environment, safety is considered as the common objective for all activities throughout the plant lifecycle. One approach to achieve plant safety is to embed safety aspects within each function and activity within such environment. One ideal way to enable safety aspects within each automated function is through modeling. This paper proposes a theoretical approach to design plant safety model as integrated with the plant lifecycle model within such environment. Object-oriented modeling approach is used to construct the plant safety model using OO CASE tool on the basis of unified modeling language (UML). Multiple views are defined for plant objects to express static, dynamic, and functional semantics of these objects. Process safety aspects are mapped to each model element and inherited from design to operation stage, as it is naturally embedded within plant's objects. By developing and realizing the plant safety model, safer plant operation can be achieved and plant safety can be assured

  2. Radiation protection aspects of design for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    The IAEA's Statute authorizes the Agency to establish safety standards to protect health and minimize danger to life and property - standards which the IAEA must use in its own operations, and which a State can apply by means of its regulatory provisions for nuclear and radiation safety. A comprehensive body of safety standards under regular review, together with the IAEA's assistance in their application, has become a key element in a global safety regime. In the mid-1990s, a major overhaul of the IAEA's safety standards programme was initiated, with a revised oversight committee structure and a systematic approach to updating the entire corpus of standards. The new standards that have resulted are of a high calibre and reflect best practices in Member States. With the assistance of the Commission on Safety Standards, the IAEA is working to promote the global acceptance and use of its safety standards. Safety standards are only effective, however, if they are properly applied in practice. The IAEA's safety services - which range in scope from engineering safety, operational safety, and radiation, transport and waste safety to regulatory matters and safety culture in organizations - assist Member States in applying the standards and appraise their effectiveness. These safety services enable valuable insights to be shared and continue to urge all Member States to make use of them. Regulating nuclear and radiation safety is a national responsibility, and many Member States have decided to adopt the IAEA's safety standards for use in their national regulations. For the Contracting Parties to the various international safety conventions, IAEA standards provide a consistent, reliable means of ensuring the effective fulfilment of obligations under the conventions. The standards are also applied by designers, manufacturers and operators around the world to enhance nuclear and radiation safety in power generation, medicine, industry, agriculture, research and education

  3. Design Features of Modern Mechanical Ventilators.

    Science.gov (United States)

    MacIntyre, Neil

    2016-12-01

    A positive-pressure breath ideally should provide a V T that is adequate for gas exchange and appropriate muscle unloading while minimizing any risk for injury or discomfort. The latest generation of ventilators uses sophisticated feedback systems to sculpt positive-pressure breaths according to patient effort and respiratory system mechanics. Currently, however, these new control strategies are not totally closed-loop systems. This is because the automatic input variables remain limited, some clinician settings are still required, and the specific features of the perfect breath design still are not entirely clear. Despite these limitations, there are some rationale for many of these newer feedback features. Copyright © 2016 Elsevier Inc. All rights reserved.

  4. FISSION 2120: a program for assessing the need for engineered safety feature grade air cleaning systems in post accident environments

    International Nuclear Information System (INIS)

    Martin, G. Jr.; Michlewicz, D.; Thomas, J.

    1979-01-01

    A computer program FISSION 2120, has been developed to evaluate the need for various engineered Safety Feature grade air cleaning systems to mitigate radiation exposures resulting from accidential releases of radioactivity. Those systems which are generally investigated include containment sprays with chemical additives, containment fan coolers with charcoal filters, and negative pressure maintenance systems for double barrier containments with either one-pass filtration or recirculation with filtration. The program can also be used to calculate the radiation doses to control room personnel. This type of analysis is directed towards the various protection aspects of the emergency ventilation system and involves the modeling of the radiological source terms and the atmospheric transport of the radioactive releases. The modeling is enhanced by the inherent capability of the program to accommodate simultaneous release of activity from several sources and to perform a dose evaluation for a wide range of the design characteristics of control room emergency air filtration systems. Use of the program has resulted in considerable savings in the time required to perform such analyses and in the selection of the most cost-effective Engineered Safety Features

  5. Design of marine structures with improved safety for environment

    International Nuclear Information System (INIS)

    Klanac, Alan; Varsta, Petri

    2011-01-01

    The paper describes a method for design of marine structures with increased safety for environment, considering also the required investment costs as well as the aspects of risk distribution onto the maritime stakeholders. Practically, the paper seeks to answer what is the optimal amount that should be invested into certain safety measure for any given vessel. Due to the uneven distribution of risk, as well as the differing impact of costs emerging from safety improvements, stakeholders experience conflicting ranking of alternatives. To solve this multi-stakeholder decision-making problem, in which each stakeholder is a decision-maker, the method applies concepts of group decision-making theory, namely the Game Theory. The method fosters axiomatic definition of the optimum solution, arguing that the solution, or the final selected design, should satisfy the non-dominance, efficiency, and fairness. These three are thoroughly discussed in terms of structural design, especially the latter. Considering the coupling of environmental risk and structural design, the method also builds on the preference structure of four maritime stakeholders: yards, owners, oil receivers and the public, who either share the risks or directly influence structural design. Method is presented on a practical study of structural design of a tanker with a crashworthy side structure that is capable of reducing the risk of collision. The outcome of this study outlines a number of possibilities for successful improvement of tanker safety that can benefit, concurrently, all maritime stakeholders.

  6. The next generation of power reactors - safety characteristics

    International Nuclear Information System (INIS)

    Modro, S.M.

    1995-01-01

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

  7. Safe-by-Design : from Safety to Responsibility

    NARCIS (Netherlands)

    van de Poel, I.R.; Robaey, Z.H.

    2017-01-01

    Safe-by-design (SbD) aims at addressing safety issues already during the R&D and design phases of new technologies. SbD has increasingly become popular in the last few years for addressing the risks of emerging technologies like nanotechnology and synthetic biology. We ask to what extent SbD

  8. Special characteristics of the safety analysis of HWRs

    International Nuclear Information System (INIS)

    Kugler, G.

    1980-01-01

    Two lectures are presented in this report. The CANDU-PHW reactor is used as a model for discussion. The first lecture describes the distinctive features of the CANDU reactor, and how they impact on reactor safety. In the second lecture the Canadian safety philosophy, the safety design objective, and other selected topics on reactor safety analysis are discussed. The material in this report was selected with a view to assisting those not familiar with the CANDU heavy water reactor design in evaluating the distinctive safety aspects of these reactors. (orig./RW)

  9. The basic discussion on nuclear power safety improvement based on nuclear equipment design

    International Nuclear Information System (INIS)

    Zhao Feiyun; Yao Yangui; Yu Hao; He Yinbiao; Gao Lei; Yao Weida

    2013-01-01

    The safety of strengthening nuclear power design was described based on nuclear equipment design after Fukushima nuclear accident. From these aspects, such as advanced standard system, advanced design method, suitable test means, consideration of beyond design basis event, and nuclear safety culture construction, the importance of nuclear safety improvement was emphatically presented. The enlightenment was given to nuclear power designer. (authors)

  10. Safety considerations in next step fusion design and beyond

    International Nuclear Information System (INIS)

    Holland, D.F.

    1990-01-01

    Recent U.S. and international design studies provide insights into the potential safety and environmental advantages of fusion as well as the development needed to realize this potential. We in the Fusion Safety Program at EG ampersand G Idaho have analyzed the Compact Ignition Tokamak (CIT), the International Thermonuclear Engineering Reactor (ITER), and the Advanced Reactor Innovative Engineering Study (ARIES). I have reviewed these three designs to determine issues related to meeting the safety and the environmental goals that guide fusion development in the U.S. The paper lists safety and environmental issues that are generic to fusion and approaches to favorably resolve each issue. The technical developments that have the highest potential of contributing to improving the safety and environmental attractiveness of fusion are identified and discussed. These developments are in the areas of low-activation materials, plasma- facing components, and plasma physics relating to off-normal plasma events and tritium burn-up. 8 refs., 7 tabs

  11. Nuclear Reactor RA Safety Report, Vol. 14, Safety protection measures

    International Nuclear Information System (INIS)

    1986-11-01

    Nuclear reactor accidents can be caused by three type of errors: failure of reactor components including (1) control and measuring instrumentation, (2) errors in operation procedure, (3) natural disasters. Safety during reactor operation are secured during its design and construction and later during operation. Both construction and administrative procedures are applied to attain safe operation. Technical safety features include fission product barriers, fuel elements cladding, primary reactor components (reactor vessel, primary cooling pipes, heat exchanger in the pump), reactor building. Safety system is the system for safe reactor shutdown and auxiliary safety system. RA reactor operating regulations and instructions are administrative acts applied to avoid possible human error caused accidents [sr

  12. User-Centered Collaborative Design and Development of an Inpatient Safety Dashboard.

    Science.gov (United States)

    Mlaver, Eli; Schnipper, Jeffrey L; Boxer, Robert B; Breuer, Dominic J; Gershanik, Esteban F; Dykes, Patricia C; Massaro, Anthony F; Benneyan, James; Bates, David W; Lehmann, Lisa S

    2017-12-01

    Patient safety remains a key concern in hospital care. This article summarizes the iterative participatory development, features, functions, and preliminary evaluation of a patient safety dashboard for interdisciplinary rounding teams on inpatient medical services. This electronic health record (EHR)-embedded dashboard collects real-time data covering 13 safety domains through web services and applies logic to generate stratified alerts with an interactive check-box function. The technological infrastructure is adaptable to other EHR environments. Surveyed users perceived the tool as highly usable and useful. Integration of the dashboard into clinical care is intended to promote communication about patient safety and facilitate identification and management of safety concerns. Copyright © 2017 The Joint Commission. All rights reserved.

  13. Key asset - inherent safety of LMFBR Pool Plant

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Sevy, R.H.; Lancet, R.T.; Mills, J.C.

    1984-04-01

    The safety approach used in the design of the Large Pool Plant emphasizes use of the intrinsic characteristics of Liquid Metal Fast Breeder Reactors to incorporate a high degree of safety in the design and reduce cost by providing simpler (more reliable) dedicated safety systems. Correspondingly, a goal was not to require the action of active systems to prevent significant core damage and/or provide large grace periods for all anticipated transients. The key safety features of the plant are presented and the analysis of representative flow and power transients are presented to show that the design goal has been satisfied

  14. Key asset--Inherent safety of LFMBR pool plant

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Lancet, R.T.; Mills, J.C.; Sevy, R.H.

    1984-01-01

    The safety approach used in the design of the Large Pool Plant emphasizes use of the intrinsic characteristics of Liquid Metal Fast Breeder Reactors to incorporate a high degree of safety in the design and reduce cost by providing simpler (more reliable) dedicated safety systems. Correspondingly, a goal was not to require the action of active systems to prevent significant core damage and/or provide large grace periods for all anticipated transients. The key safety features of the plant are presented and the analysis of representative flow and power transients are presented to show that the design goal has been satisfied

  15. Development and design of a computer-assisted information management system for radiation safety management at the University of Washington

    International Nuclear Information System (INIS)

    Riches, C.G.; Riordan, F.J.; Robb, D.; Grieb, C.; Pence, G.; O'Brien, M.J.

    1984-01-01

    The Radiation Safety Office (RSO) at the University of Washington (UW) found that it needed a computerized information system to help manage the campus radiation safety program and to help provide the records necessary to show compliance with regulations and license requirements. The John L. Locke Computer Center at the UW had just developed the GLAMOR system to aid information entry and query for their computer when the RSO turned to them for assistance. The module that was developed provided a mechanism for controlling and monitoring radioactive materials on campus. This became one part of a multi-faceted system that registers users, employees, sealed sources and radiation-producing machines. The system is designed to be interactive, for immediate information recall, and powerful enough to provide routine and special reports on compliance status. The RSO information system is designed to be flexible and can easily incorporate additional features. Some future features include an interactive SNM control program, an interface to the information system currently being developed for the occupational safety and health program and an interface to the database provided by the commercial film badge service used by the University. Development of this program lead the RSO to appreciate the usefulness of having health physics professionals on the staff who were also knowledgeable about computers and who could develop programs and reports necessary to their activities

  16. A cost-effective methodology to internalize nuclear safety in nuclear reactor conceptual design

    International Nuclear Information System (INIS)

    Gimenez, M.; Grinblat, P.; Schlamp, M.

    2003-01-01

    A new methodology to perform nuclear reactor design, balancing safety and economics at the conceptual engineering stage, is presented in this work. The goal of this integral methodology is to take into account safety aspects in an optimization design process where the design variables are balanced in order to obtain a better figure of merit related with reactor economic performance. Design parameter effects on characteristic or critical safety variables, chosen from reactor behavior during accidents (safety performance indicators), are synthesized on Design Maps. These maps allow one to compare the safety indicator with limits, which are determined by design criteria or regulations, and to transfer these restrictions to the design parameters. In this way, reactor dynamic response and other safety aspects are integrated in a global optimization process, by means of additional rules to the neutronic, thermal-hydraulic, and mechanical calculations. An application of the methodology, implemented in Integrated Reactor Evaluation Program 3 (IREP3) code, to optimize safety systems of CAREM prototype is presented. It consists in balancing the designs of the Emergency Injection System (EIS), the Residual Heat Removal System (RHRS), the primary circuit water inventory and the containment height, to cope with loss of coolant and loss of heat sink (LOHS) accidental sequences, taking into account cost and reactor performance. This methodology turns out to be promising to internalize cost-efficiently safety issues. It also allows one to evaluate the incremental costs of implementing higher safety levels

  17. Advanced power reactors with improved safety characteristics

    International Nuclear Information System (INIS)

    Birkhofer, A.

    1994-01-01

    The primary objective of nuclear safety is the protection of individuals, society and environment against radiological hazards from accidental releases of radioactive materials contained in nuclear reactors. Hereto, these materials are enclosed by several successive barriers and the barriers protected against mishaps and accidents by a multi-level system of safety precautions. The evolution of reactor technology continuously improves this concept and its implementation. At a world-wide scale, several advanced reactor concepts are currently being considered, some of them already at a design stage. Essential safety objectives include both further strengthening the prevention of accidents and improving the containment of fission products should an accident occur. The proposed solutions differ considerably with regard to technical principles, plant size and time scales considered for industrial application. Two typical approaches can be distinguished: The first approach basically aims at an evolution of power reactors currently in use, taking into account the findings from safety research and from operation of current plants. This approach makes maximum use of proven technology and operating experience but may nevertheless include new safety features. The corresponding designs are often termed 'large evolutionary'. The second approach consists in more fundamental changes compared to present designs, often with strong emphasis on specific passive features protecting the fuel and fuel cladding barriers. Owing to the nature and capability of those passive features such 'innovative designs' are mostly smaller in power output. The paper describes the basic objectives of such developments and illustrates important technical concepts focusing on next generation plants, i.e. designs to be available for industrial application until the end of this decade. 1 tab. (author)

  18. Conceptual design of safety instrumentation for PFBR

    International Nuclear Information System (INIS)

    Muralikrishna, G.; Seshadri, U.; Raghavan, K.

    1996-01-01

    Instrumentation systems enable monitoring of the process which in turn enables control and shutdown of the process as per the requirements. Safety Instrumentation due to its vital importance has a stringent role and this needs to be designed methodically. This paper presents the details of the conceptual design for PFBR. (author). 4 figs, 3 tabs

  19. Ageing study of the engineered safety features actuation system of the Loviisa NPP

    International Nuclear Information System (INIS)

    Simola, K.; Maskuniitty, M.

    1995-06-01

    An ageing study of the engineered safety features actuation system of the Loviisa nuclear power plant has been performed. The operating experience, including failure and maintenance histories of analog measuring devices, logics for safety signal formation and individual control electronics of pumps and valves, has been collected and analysed. The safety importance of system components has been studied with a fault tree analysis of a selected safety function. Based on the results of the analysis of operating experiences and the fault tree analysis, some components were selected for deeper analyses. According to the operating experience, the amount of failures in the Loviisa plant safety system has been low and no increasing trend in the failure history can yet be observed. Only a few failures had prohibited the propagation of the safety signal, mostly the failures have caused a false alarm. The failures reported have concerned mainly limit signal units, transmitters, and priority units. According to the fault tree analysis of one safety function, the most important components of this subsystem are individual control units and pulse/DC converters. Failure modes and effect analyses were performed for priority and individual control unit, limit signal unit and comparator and pulse/DC converter in order to identify the critical failure modes of these devices. (orig.) (15 refs., 26 figs., 9 tabs.)

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

    International Nuclear Information System (INIS)

    Hongbin Zhang; Haihua Zhao; Vincent Mousseau

    2008-01-01

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

  1. Risk-informed approach for safety, safeguards, and security (3S) by design

    International Nuclear Information System (INIS)

    Suzuki, Mitsutoshi; Burr, Tom; Howell, John

    2011-01-01

    Over several decades the nuclear energy society worldwide has developed safety assessment methodology based on probabilistic risk analysis for incorporating its benefit into design and accident prevention for nuclear reactors. Although safeguards and security communities have different histories and technical aspects compared to safety, risk assessment as a supplement to their current requirements could be developed to promote synergism between Safety, Safeguards, and Security (3S) and to install effective countermeasures in the design of complex nuclear fuel cycle facilities. Since the 3S initiative was raised by G8 countries at Hokkaido Toyako-Summit in 2008, one approach to developing synergism in a 3S By Design (3SBD) process has been the application of risk-oriented assessment methodology. In the existing regulations of safeguards and security, a risk notion has already been considered for inherent threat and hazard recognition. To integrate existing metrics into a risk-oriented approach, several mathematical methods have already been surveyed, with attention to the scarcity of intentional acts in the case of safeguards and the sparseness of actual event data. A two-dimensional probability distribution composed of measurement error and incidence probabilities has been proposed to formalize inherent difficulties in the International Atomic Energy Agency (IAEA) safeguards criteria. In particular, the incidence probability that is difficult to estimate has been explained using a Markov model and game theory. In this work, a feasibility study of 3SBD is performed for an aqueous reprocessing process, and synergetic countermeasures are presented for preliminary demonstration of 3SBD. Although differences and conflicts between individual 'S' communities exist, the integrated approach would be valuable for optimization and balance between the 3S design features as well as for effective and efficient implementation under existing regulation frameworks. In addition

  2. Designing a Safety Reporting Smartphone Application to Improve Patient Safety After Total Hip Arthroplasty.

    Science.gov (United States)

    Krumsvik, Ole Andreas; Babic, Ankica

    2017-01-01

    This paper presents a safety reporting smartphone application which is expected to reduce the occurrence of postoperative adverse events after total hip arthroplasty (THA). A user-centered design approach was utilized to facilitate optimal user experience. Two main implemented functionalities capture patient pain levels and well-being, the two dimensions of patient status that are intuitive and commonly checked. For these and other functionalities, mobile technology could enable timely safety reporting and collection of patient data out of a hospital setting. The HCI expert, and healthcare professionals from the Haukeland University Hospital in Bergen have assessed the design with respect to the interaction flow, information content, and self-reporting functionalities. They have found it to be practical, intuitive, sufficient and simple for users. Patient self-reporting could help recognizing safety issues and adverse events.

  3. Evaluating public awareness of new currency design features

    Science.gov (United States)

    DiNunzio, Lisa; Church, Sara E.

    2002-04-01

    One of the goals of the 1996 series design was to integrate highly recognizable features that enable the general public to more easily distinguish counterfeit from genuine notes, thereby reducing the chance of counterfeit notes being passed. The purpose of this study is to evaluate how knowledgeable the public is concerning the new currency, to identify the channels through which the public learns about new currency design, and to assess the usefulness of the new currency's authentication features. Also, the study will serve as a baseline measurement for future design studies and in comparative analysis with other countries. The results of the qualitative research will be described in the following sections of this paper. The quantitative research is scheduled to begin in February 2002, at the same time as the Netherlands' opinion poll of the Euro and NLG-notes in an effort to compare results.

  4. Safety margin improvement by adopting the feature of interleaving in 700 MWe PHWR

    International Nuclear Information System (INIS)

    Kumar, Nrependra; Yadav, S.K.; Khan, T.A.; Dixit, A.; Singhal, Mukesh; Nair, Suma R.

    2015-01-01

    Indian Pressurised Heavy Water Reactors (IPHWRs) of 700 MWe are under construction at Kakrapar Atomic Power Project -3,4 and Rajasthan Atomic Power Project-7,8. These units have enhanced safety features with respect to standard IPHWRs. One of the enhanced features is interleaving of feeders/channels. In interleaved feeder configuration, each header located at either end of reactor gets connected to one quarter of core channels, which are uniformly distributed. The core is divided into two loops with feeder connected in interleaved fashioned. In this paper a comparative study has been performed between the two cases: 1) The core splits in two vertical halves and each vertical half is a loop of PHT (TAPS-3 and 4 Type configuration). 2) The core is divided into two loops with feeders/ channels connected in interleaved fashioned (700 MWe Configuration). LOCA studies have been performed for 700 MWe PHWR considering interleaving of feeders configuration using in-house developed computer code ATMIKA and 3-D neutron kinetics code IQS-3D. The issue of interleaving is closely linked to an inherent reactivity characteristic of PHWR reactors (viz., positive void reactivity coefficient) which leads to a power increase following a Large LOCA. In 700 MWe PHWR with intent to improve the safety margin, adopted the feature of interleaving of feeders which causes in reduction in the magnitude of void coefficient and results in reduction of peak power during LBLOCA. The systematic LBLOCA study demonstrates that interleaved configuration of feeder/channels of two loops has higher safety margins (i.e. with respect to peak power, prompt-criticality margin, adiabatic heat deposition on the fuel pins, sheath temperature excursion and clad oxidation) with regard to the effectiveness of shutdown system. (author)

  5. U.S. ALMR safety approach and licensing status

    International Nuclear Information System (INIS)

    Herczeg, J.W.; Hardy, R.W.; Gyorey, G.L.

    1992-01-01

    The Advanced Liquid Metal Cooled Reactor (ALMR) in the United States is based on the Power Reactor Innovative Small Module (PRISM) concept originated by the General Electric Company (GE). This concept features a compact modular system suitable for factory fabrication, and a high degree of passive and natural safety characteristics. The safety approach emphasizes accident prevention, backed up by accident mitigation. First-round safety evaluations by U.S. regulators have found that the design provides passive, natural, and other desirable features enhancing the safety of the power plant. A Preapplication Safety Evaluation Report (PSER) from the U.S. Nuclear Regulatory Commission (NRC) is anticipated in early 1993. (author)

  6. Programmable electronic safety systems

    International Nuclear Information System (INIS)

    Parry, R.R.

    1993-01-01

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

  7. The safety relief valve handbook design and use of process safety valves to ASME and International codes and standards

    CERN Document Server

    Hellemans, Marc

    2009-01-01

    The Safety Valve Handbook is a professional reference for design, process, instrumentation, plant and maintenance engineers who work with fluid flow and transportation systems in the process industries, which covers the chemical, oil and gas, water, paper and pulp, food and bio products and energy sectors. It meets the need of engineers who have responsibilities for specifying, installing, inspecting or maintaining safety valves and flow control systems. It will also be an important reference for process safety and loss prevention engineers, environmental engineers, and plant and process designers who need to understand the operation of safety valves in a wider equipment or plant design context. . No other publication is dedicated to safety valves or to the extensive codes and standards that govern their installation and use. A single source means users save time in searching for specific information about safety valves. . The Safety Valve Handbook contains all of the vital technical and standards informat...

  8. Safety design of Pb-Bi-cooled direct contact boiling water fast reactor (PBWFR)

    International Nuclear Information System (INIS)

    Takahashi, Minoru; Uchida, Shoji; Yamada, Yumi; Koyama, Kazuya

    2008-01-01

    In Pb-Bi-cooled direct contact boiling water small fast reactor (PBWFR), steam is generated by direct contact of feedwater with primary Pb-Bi coolant above the core, and Pb-Bi coolant is circulated by steam lift pump in chimneys. Safety design has been developed to show safety features of PBWFR. Negative void reactivity is inserted even if whole of the core and upper plenum are voided hypothetically by steam intrusion from above. The control rod ejection due to coolant pressure is prevented using in-vessel type control rod driving mechanism. At coolant leak from reactor vessel and feedwater pipes, Pb-Bi coolant level in the reactor vessel required for decay heat removal is kept using closed guard vessel. Dual pipes for feedwater are employed to avoid leak of water. Although there is no concern of loss of flow accident due to primary pump trip, feedwater pump trip initiates loss of coolant flow (LOF). Injection of high pressure water slows down the flow coast down of feedwater at the LOF event. The unprotected loss of flow and heat sink (ATWS) has been evaluated, which shows that the fuel temperatures are kept lower than the safety limits. (author)

  9. ARIES-RS safety design and analysis

    International Nuclear Information System (INIS)

    Steiner, D.; El-Guebaly, L.; Herring, S.; Khater, H.; Mogahed, E.; Thayer, R.; Tillack, M.S.

    1997-01-01

    The ARIES-RS safety design and analysis focused on achieving two objectives: (1) The avoidance of sheltering or evacuation in the event of an accident; and (2) the generation of only low-level waste, no greater than Class C. The ARIES-RS baseline design employs V-4Cr-4Ti as the blanket structural material and a low activation ferritic steel in the reflector and shield. In the event of a LOCA, the baseline design first wall maximum temperature falls in the range of 1100-1200 C. For this temperature range, the hazard assessment indicates that the dose at the site boundary will be less than 1 rem per year. Thus, no sheltering or evacuation would be required in the event of a LOCA. Although the baseline design satisfies the first safety objective noted above, a first wall maximum temperature of ∝1100-1200 C would likely compromise the integrity of the vanadium blanket structure and would require blanket replacement following such a temperature excursion. To avoid this situation, a modified blanket design incorporating supplemental heat removal is also proposed. Preliminary analysis of this modified design suggests that the first wall maximum temperature can be kept below the temperature range of concern, ∝1000-1100 C, in the event of a LOCA. When the ferritic steel used in the reflector and shield is one reduced in Ir and Ag impurities, all in-vessel components qualify for near-surface shallow land burial as Class C low-level waste. (orig.)

  10. Predicting the effectiveness of road safety campaigns through alternative research designs.

    Science.gov (United States)

    Adamos, Giannis; Nathanail, Eftihia

    2016-12-01

    A large number of road safety communication campaigns have been designed and implemented in the recent years; however their explicit impact on driving behavior and road accident rates has been estimated in a rather low proportion. Based on the findings of the evaluation of three road safety communication campaigns addressing the issues of drinking and driving, seat belt usage, and driving fatigue, this paper applies different types of research designs (i.e., experimental, quasi-experimental, and non-experimental designs), when estimating the effectiveness of road safety campaigns, implements a cross-design assessment, and conducts a cross-campaign evaluation. An integrated evaluation plan was developed, taking into account the structure of evaluation questions, the definition of measurable variables, the separation of the target audience into intervention (exposed to the campaign) and control (not exposed to the campaign) groups, the selection of alternative research designs, and the appropriate data collection methods and techniques. Evaluating the implementation of different research designs in estimating the effectiveness of road safety campaigns, results showed that the separate pre-post samples design demonstrated better predictability than other designs, especially in data obtained from the intervention group after the realization of the campaign. The more constructs that were added to the independent variables, the higher the values of the predictability were. The construct that most affects behavior is intention, whereas the rest of the constructs have a lower impact on behavior. This is particularly significant in the Health Belief Model (HBM). On the other hand, behavioral beliefs, normative beliefs, and descriptive norms, are significant parameters for predicting intention according to the Theory of Planned Behavior (TPB). The theoretical and applied implications of alternative research designs and their applicability in the evaluation of road safety

  11. Design concepts and safety concerns of the small and medium size reactors (SMR)

    International Nuclear Information System (INIS)

    Seul, Kwang Won; Lee, Jae Hun; Kim, Hho Jung

    1998-01-01

    The small and medium size reactors (SMR) and interface facilities such as desalination plant are expected to be located near the population area because of restrictions in transporting the plant products such as fresh water to long distance area. To protect the public around the plant facility from the possible release of radioactive materials, the design development of the SMR is focusing on an enhancement of the safety and reliability as well as the economics. In this study, the major safety concepts of the SMR designs significantly different from the current PWR designs are investigated and the safety concerns applicable to the integrated SMR design of Korea (called SMART), were identified. Those safety issues include the use of proven technology, application of strengthening defense in depth, event categorization and selection, simplification of emergency planning, determination of accident source terms and so on. The efforts to resolve the safety concerns in the design stage will provide an improvement of the safety of the SMART design

  12. Final safety evaluation report related to the certification of the System 80{sup +} design (Docket No. 52-002). Volume 1, Chapters 1--14

    Energy Technology Data Exchange (ETDEWEB)

    1994-08-01

    This final safety evaluation report (FSER) documents the technical review of the System 80+ standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the System 80+ design was submitted by Combustion Engineering, Inc., now Asea Brown Boveri-Combustion Engineering (ABB-CE) as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. System 80+ is a pressurized water reactor with a rated power of 3914 megawatts thermal (MWt) and a design power of 3992 MWt at which accidents are analyzed. Many features of the System 80+ are similar to those of Abb-CE`s System 80 design from which it evolved. Unique features of the System 80+ design included: a large spherical, steel containment; an in-containment refueling water storage tank; a reactor cavity flooding system, hydrogen ignitors, and a safety depressurization system for severe accident mitigation; a combustion gas turbine for an alternate ac source; and an advanced digitally based control room. On the basis of its evaluation and independent analyses, the NRC staff concludes that ABB-CE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the System 80+ standard design. This document, Volume 1, contains Chapters 1 through 14 of this report.

  13. Safety in waste management plants: An Indian perspective

    International Nuclear Information System (INIS)

    Shekhar, P.; Ozarde, P.D.; Gandhi, P.M.

    2000-01-01

    Assurance of safety of public and plant workers and protection of the environment are prime objectives in the design and construction of Waste Management Plants. In India, waste management principles and strategies have been evolved in accordance with national and international regulations and standards for radiation protection. The regulations governing radiation protection have a far-reaching impact on the management of the radioactive waste. The wastes arise at each stages of the fuel cycle with varying chemical nature, generation rate and specific activity levels depending upon the type of the facility. Segregation of waste based on its chemical nature and specific activity levels is an essential feature, as its aids in selection of treatment and conditioning process. Selection of the process, equipment and materials in the plant, are governed by safety consideration alongside factors like efficiency and simplicity. The plant design considerations like physical separation, general arrangement, ventilation zoning, access control, remote handling, process piping routing, decontamination etc. have major role in realizing waste safety. Stringent quality control measures during all stages of construction have helped in achieving the design intended safety. These aspects together with operating experience gained form basis for the improved safety features in the design and construction of waste management plants. The comprehensive safety is derived from adoption of waste management strategies and appropriate plant design considerations. The paper briefly brings safety in waste management programme in India, in its current perspective. (author)

  14. Development of safety principles for the design of future nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-06-01

    The main purpose of this TECDOC is to propose updates to existing safety principles which could be used as a basis for developing safety principles for the design of future NPPs. Accordingly, this document is intended to be useful to reactor designers, owners, operators, researchers and regulators. It is also expected that this document can contribute to international harmonization of safety approaches, and that it will help ensure that future reactors will be designed worldwide to a high standard of safety. As such, these proposed updates are intended to provide general guidance which, if carefully and properly implemented, will result in reactor designs with enhanced safety characteristics beyond those currently in operation. This enhancement results from the fact that the proposals are derived from the lessons learned from more recent operational experience, R and D, design, testing, and analysis developed over the past decade or so, as well as from attempts to reflect the current trends in reactor design, such as the introduction of new technologies. 8 refs, 3 figs.

  15. Development of safety principles for the design of future nuclear power plants

    International Nuclear Information System (INIS)

    1995-06-01

    The main purpose of this TECDOC is to propose updates to existing safety principles which could be used as a basis for developing safety principles for the design of future NPPs. Accordingly, this document is intended to be useful to reactor designers, owners, operators, researchers and regulators. It is also expected that this document can contribute to international harmonization of safety approaches, and that it will help ensure that future reactors will be designed worldwide to a high standard of safety. As such, these proposed updates are intended to provide general guidance which, if carefully and properly implemented, will result in reactor designs with enhanced safety characteristics beyond those currently in operation. This enhancement results from the fact that the proposals are derived from the lessons learned from more recent operational experience, R and D, design, testing, and analysis developed over the past decade or so, as well as from attempts to reflect the current trends in reactor design, such as the introduction of new technologies. 8 refs, 3 figs

  16. Main circulator design features for HTR 100, HTR 500 and space heating plants

    International Nuclear Information System (INIS)

    Engel, J.; Glass, D.

    1988-01-01

    All design alternatives for modern high-temperature reactors have a common circulator concept: It is based on a vertical shaft design with a flying impeller. The circulators are equipped with active magnetic bearings and are driven by induction motors connected to variable-speed static converters. Due to their multiple functions during normal reactor operation and under accident conditions, extremely high requirements are made to safety-relevant circulators, since with the reactor pressurized as well as under depressurized conditions specified delivery heads and flow rates have to be ensured. The use of active magnetic bearings permits to obtain maintenance-free operation and functional safety to an extent which had not been achieved before. Magnetic bearings are therefore provided for the total range including primary gas circulators of a drive power of several MW as well as circulators for helium loops of reactor auxiliary systems. The essential feature for using active magnetic bearings is the retainer bearing technology, preventing contact between rotor and static circulator parts upon unintended deenergisation of the magnets. Results of current experiments are reported. Another aspect to be considered for reliable long-term operation for several decades is the effect of rotor dynamics. The various natural frequencies resulting from torsion and bending modes in view of a drive by a frequency-controlled induction motor have to be considered as well as the specific characteristics of the active magnetic bearings. Special attention has to be directed to the internal cooling loop so as to ensure that reactor temperature excursions in the event of deviation from normal operation can be overcome without damage. For circulator components exposed to temperature fields the design characteristics are determined by combining experimental and analytical methods. The coordination of all component parts is currently being optimized on a prototype circulator whose detailed

  17. Nuclear criticality safety program at the Fuel Cycle Facility

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  18. A proposed approach for enhancing design safety assurance of future plants

    International Nuclear Information System (INIS)

    Oh, Kyu Myeng; Ahn, Sang Kyu; Lee, Chang Ju; Kim, Inn Seock

    2010-01-01

    This paper provides various insights from a detailed review of deterministic approaches typically applied to ensure design safety of nuclear power plants (NPPs) and risk-informed approaches proposed to evaluate safety of advanced reactors such as Generation IV reactors. Also considered herein are the risk-informed safety analysis (RISA) methodology suggested by Westinghouse as a means to improve the conventional accident analysis, together with the Technology Neutral Framework recently suggested by the U.S. NRC for safety evaluation of future plants. These insights from the comparative review of deterministic and risk-informed approaches could be used in further enhancing the methodology for design safety assurance of future plants

  19. Molten salt reactors - safety options galore

    International Nuclear Information System (INIS)

    Gat, U.; Dodds, H.L.

    1997-01-01

    Safety features and attributes of molten salt reactors (MSR) are described. The unique features of fluid fuel reactors of on-line continuous processing and the ability for so-called external cooling result in simple and safe designs with low excess reactivity, low fission product inventory, and small source term. These, in turn, make a criticality accident unlikely and reduce the severity of a loss of coolant to where they are no longer severe accidents. A melt down is not an accident for a reactor that uses molten fuel. The molten salts are stable, non-reactive and efficient heat transfer media that operate at high temperatures at low pressures and are highly compatible with selected structural materials. All these features reduce the accident plethora. Freeze valves can be used for added safety. An ultimate safe reactor (U.S.R) is described with safety features that are passive, inherent and non-tamperable (PINT)

  20. Operational safety at the FFTF

    International Nuclear Information System (INIS)

    Baird, Q.L.; Hagan, J.W.; Seeman, S.E.; Baker, S.M.

    1981-02-01

    An extensive operational nuclear safety program has been an integral part of the design, startup, and initial operating phases of the Fast Flux Test Facility (FFTF). During the design and construction of the facility, a program of independent safety overviews and analyses assured the provision of responsible safety margins within the plant, protective systems, and engineered safety features for protection of the public, operating staff, and the facility. The program is continuing through surveillance of operations to verify continued adherence to the established operating envelope and for timely identification of any trends potentially adverse to those margins. Experience from operation of FFTF is being utilized in the development of enhanced operational nuclear safety aids for application in follow-on breeder reactor power systems. The commendable plant and personnel safety experiences of FFTF through its startup and ascension to full power demonstrate the overall effectiveness of the FFTF operational nuclear safety program

  1. CANDU 9 Design improvements based on experience feedback

    International Nuclear Information System (INIS)

    Yu, S. K. W.; Bonechi, M.; Snell, V. G.

    2000-01-01

    An evolutionary approach utilizing advance technologies has been implenented for the enhancement introduced in the CANDU 9 Nuclear Power Plant (NPP) design. The design of these systems and associated equipment has also benfited from experience feedback from operating CANDU stations and from including advanced products from CANDU engineering and research programs. This paper highlights the design features that contribute to the safety improvements of the CANDU 9 design, summarizes the analysis results which demonstrate the improved performance and also emphasizes design features which reduce operation and maintenance (Q and M) costs. The safety design features highlighted include the increased use of passive devices and heat sinks to achieve extensive system simplification; this also improves reliability and reduces maintenance workloads. System features that contribute to improved operability are also described. The CANDU 9 Control Center provides plant staff with enhanced operating, maintenance and diagnostics features which significantly improve operability, testing and maintainability due to the integration of human factors engineering with a systematic design process. (author)

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

    International Nuclear Information System (INIS)

    Shoji Katanishi; Kazuhiko Kunitomi; Shusaku Shiozawa

    2002-01-01

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

  3. Safety and security aspects in design of digital safety I and C in nuclear power plants

    International Nuclear Information System (INIS)

    Ding, Yongjian; Waedt, Karl

    2016-01-01

    The paper describes a safety objective oriented systematic design approach of digital (computerized) safety I and C in modern nuclear power plants which considers the plant safety requirements as well as cybersecurity needs. The defence in depth philosophy is applied by using different defence lines in the I and C architecture and protection zones in the plant IT environment.

  4. Safety and security aspects in design of digital safety I and C in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Yongjian [University of Applied Sciences Magdeburg-Stendal, Magdeburg (Germany). Inst. of Electrical Engineering; Waedt, Karl [Areva GmbH, Erlangen (Germany). PEAS-G

    2016-05-15

    The paper describes a safety objective oriented systematic design approach of digital (computerized) safety I and C in modern nuclear power plants which considers the plant safety requirements as well as cybersecurity needs. The defence in depth philosophy is applied by using different defence lines in the I and C architecture and protection zones in the plant IT environment.

  5. Design and safety considerations for the 10 MW(t) multipurpose TRIGA reactor in Thailand

    International Nuclear Information System (INIS)

    Razvi, J.; Bolin, J.M.; Saurwein, J.J.; Whittemore, W.L.; Proongmuang, S.

    1999-01-01

    containment structure is not required. Thus the reactor will be housed in a confinement building. The basic design of the reactor, reactor structure, auxiliary systems, reactor instrumentation and control systems and other balance of plant systems have been completed and detailed design is underway. The Preliminary Safety Analysis Report (PSAR) has been completed and submitted to OAEP for approval and issuance of a Construction Permit. The PSAR has been reviewed by OAEP, its consultants as well as the IAEA. Fuel loading and commissioning is expected before the end of 2002. This paper describes the basic design features of the new reactor, including key features of the reactor fuel, core and related structures. The results of reactor performance and safety analyses performed in support of the PSAR are described. (author)

  6. Design philosophy of PFBR shutdown systems

    International Nuclear Information System (INIS)

    Rajan Babu, V.; Vijayashree, R.; Govindarajan, S.; Vaidyanathan, G.; Muralikrishna, G.; Shanmugam, T.K.; Chetal, S.C.; Raghavan, K.; Bhoje, S.B.

    1996-01-01

    This paper presents the overall design philosophy of shutdown system of 500 MWe Prototype Fast Breeder Reactor (PFBR). It discusses design criteria, parameters calling for safety action, different safety actions and the concepts conceived for shutdown systems. In tune with the philosophy of defence-in-depth, additional passive shutdown features, viz., Self Actuating Device (SADE) and Curie Point Magnetic (CPM) switch and protective feature like absorber rod Stroke Limiting Device (SLD) are contemplated. It also discusses about suitability of Gas Expansion Module (GEM) as one of the safety devices in PFBR. (author). 3 refs, 3 figs, 1 tab

  7. Meeting up-to-date safety requirements in the Russian NPP projects

    International Nuclear Information System (INIS)

    Tepkyan, G. O.; Yashkin, A. V.

    2014-01-01

    Safety features in Russian NPP designs are implemented by the combination of active and passive safety systems • Russian NPP designs are in compliance with up-to-date international and European safety requirements and refer to Generation III+ • Russian state-of-the-art designs have already implemented some design solutions, which take into account “post-Fukushima” requirements. Russian NPP design principles have been approved during the European discussions in spring 2012, including the IAEA extraordinary session addressed to Fukushima NPP accident

  8. Safety regulation for the design approval of special form radioactive sources

    International Nuclear Information System (INIS)

    Cho, Woon-Kap

    2009-01-01

    Several kinds of special form radioactive sources for industrial, medical applications are being produced in Korea. Special form radioactive sources should meet strict safety requirements specified in the domestic safety regulations and the design of the sources should be certified by the regulatory authority, the Ministry of Education, Science and Technology (MEST). Several safety tests such as impact, percussion, heating, and leak tests are performed on the sources according to the domestic regulations and the international safety standards such as ANSI N542-1977 and ISO 2919-1999(E). As a regulatory expert body, Korea Institute of Nuclear Safety (KINS) assesses various types of application documents, such as safety analysis report, quality assurance program, and other documents evidencing fulfillment of requirements for design approval of the special form radioactive sources, submitted by a legal person who intends to produce special form radioactive sources and then reports the assessment result to MEST. A design approval certificate is issued to the applicant by MEST on the basis of a technical evaluation report presented by KINS.

  9. ASIC-based design of NMR system health monitor for mission/safety-critical applications.

    Science.gov (United States)

    Balasubramanian, P

    2016-01-01

    N-modular redundancy (NMR) is a generic fault tolerance scheme that is widely used in safety-critical circuit/system designs to guarantee the correct operation with enhanced reliability. In passive NMR, at least a majority (N + 1)/2 out of N function modules is expected to operate correctly at any time, where N is odd. Apart from a conventional realization of the NMR system, it would be useful to provide a concurrent indication of the system's health so that an appropriate remedial action may be initiated depending upon an application's safety criticality. In this context, this article presents the novel design of a generic NMR system health monitor which features: (i) early fault warning logic, that is activated upon the production of a conflicting result by even one output of any arbitrary function module, and (ii) error signalling logic, which signals an error when the number of faulty function modules unfortunately attains a majority and the system outputs may no more be reliable. Two sample implementations of NMR systems viz. triple modular redundancy and quintuple modular redundancy with the proposed system health monitoring are presented in this work, with a 4-bit ALU used for the function modules. The simulations are performed using a 32/28 nm CMOS process technology.

  10. Safety design guides for containment extension for CANDU 9

    International Nuclear Information System (INIS)

    Lee, Duk Su; Chang, Woo Hyun; Lee, Nam Young; A. C. D. Wright

    1996-03-01

    This safety design guide for containment extension describes the containment isolation philosophy and containment extension requirements. The metal extensions and components falling within the scope of ASME Section III are classified in accordance with the CAN/CSA-N285.0 and CAN/CSA-N285.3. The special consideration for the leak monitoring capability, seismic qualification and inspection requirements for containment extensions, etc., are defined in this design guide. In addition, the containment isolation systems are defined and summarized schematically in appendix A. The change status of the regulatory requirements, code and standards should be traced and this safety design guide shall be updated accordingly. (Author) .new

  11. Multi-dimensional database design and implementation of dam safety monitoring system

    Directory of Open Access Journals (Sweden)

    Zhao Erfeng

    2008-09-01

    Full Text Available To improve the effectiveness of dam safety monitoring database systems, the development process of a multi-dimensional conceptual data model was analyzed and a logic design was achieved in multi-dimensional database mode. The optimal data model was confirmed by identifying data objects, defining relations and reviewing entities. The conversion of relations among entities to external keys and entities and physical attributes to tables and fields was interpreted completely. On this basis, a multi-dimensional database that reflects the management and analysis of a dam safety monitoring system on monitoring data information has been established, for which factual tables and dimensional tables have been designed. Finally, based on service design and user interface design, the dam safety monitoring system has been developed with Delphi as the development tool. This development project shows that the multi-dimensional database can simplify the development process and minimize hidden dangers in the database structure design. It is superior to other dam safety monitoring system development models and can provide a new research direction for system developers.

  12. Nuclear safety as applied to space power reactor systems

    International Nuclear Information System (INIS)

    Cummings, G.E.

    1987-01-01

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

  13. Design of agricultural product quality safety retrospective supervision system of Jiangsu province

    Science.gov (United States)

    Wang, Kun

    2017-08-01

    In store and supermarkets to consumers can trace back agricultural products through the electronic province card to query their origin, planting, processing, packaging, testing and other important information and found that the problems. Quality and safety issues can identify the responsibility of the problem. This paper designs a retroactive supervision system for the quality and safety of agricultural products in Jiangsu Province. Based on the analysis of agricultural production and business process, the goal of Jiangsu agricultural product quality safety traceability system construction is established, and the specific functional requirements and non-functioning requirements of the retroactive system are analyzed, and the target is specified for the specific construction of the retroactive system. The design of the quality and safety traceability system in Jiangsu province contains the design of the overall design, the trace code design and the system function module.

  14. Planning and architectural safety considerations in designing nuclear power plants

    International Nuclear Information System (INIS)

    Konsowa, Ahmed A.

    2009-01-01

    To achieve optimum safety and to avoid possible hazards in nuclear power plants, considering architectural design fundamentals and all operating precautions is mandatory. There are some planning and architectural precautions should be considered to achieve a high quality design and construction of nuclear power plant with optimum safety. This paper highlights predicted hazards like fire, terrorism, aircraft crash attacks, adversaries, intruders, and earthquakes, proposing protective actions against these hazards that vary from preventing danger to evacuating and sheltering people in-place. For instance; using safeguards program to protect against sabotage, theft, and diversion. Also, site and building well design focusing on escape pathways, emergency exits, and evacuation zones, and the safety procedures such as; evacuation exercises and sheltering processes according to different emergency classifications. In addition, this paper mentions some important codes and regulations that control nuclear power plants design, and assessment methods that evaluate probable risks. (author)

  15. Safety and environmental aspects in LNG carrier design

    International Nuclear Information System (INIS)

    Takashi Yoneyama

    1997-01-01

    'Safety and Reliability' has been and will continue to be a key phr ase in marine transportation of LNG. Mitsui Engineering and Shipbuilding Co.,Ltd. has utilized its all expertise and state of art technologies to realize this objective, resulting in exceptionally successful operations of LNG carrier built by the Co. In line with growing global concern about environmental issues, we need to pay more attention to the environmental aspects of the design and construction of LNG carriers. Accordingly, in this paper, we present some topics related safety and environmental concerns which need to be taken into consideration in LNG carriers design and construction. (Author). 7 figs

  16. Safety and environmental aspects in LNG carrier design

    Energy Technology Data Exchange (ETDEWEB)

    Yoneyama, Takashi [Mitsui Shipbuilding and Engineering Co. Ltd., Tokyo (Japan)

    1997-06-01

    `Safety and Reliability` has been and will continue to be a key phr ase in marine transportation of LNG. Mitsui Engineering and Shipbuilding Co.,Ltd. has utilized its all expertise and state of art technologies to realize this objective, resulting in exceptionally successful operations of LNG carrier built by the Co. In line with growing global concern about environmental issues, we need to pay more attention to the environmental aspects of the design and construction of LNG carriers. Accordingly, in this paper, we present some topics related safety and environmental concerns which need to be taken into consideration in LNG carriers design and construction. (Author). 7 figs.

  17. Research on conceptual design of simplified nuclear safety instrument and control system

    International Nuclear Information System (INIS)

    Huang Jie

    2015-01-01

    The Nuclear safety instrument and control system is directly related to the safety of the reactor. So redundant and diversity design is used to ensure the system's security and reliability. This make the traditional safety system large, more cabinets and wiring complexity. To solve these problem, we can adopt new technology to make the design more simple. The simplify conceptual design can make the system less cabinets, less wiring, but high security, strong reliability. (author)

  18. Code on the safety of nuclear power plants: Design

    International Nuclear Information System (INIS)

    1988-01-01

    This Code is a compilation of nuclear safety principles aimed at defining the essential requirements necessary to ensure nuclear safety. These requirements are applicable to structures, systems and components, and procedures important to safety in nuclear power plants embodying thermal neutron reactors, with emphasis on what safety requirements shall be met rather than on specifying how these requirements can be met. It forms part of the Agency's programme for establishing Codes and Safety Guides relating to land based stationary thermal neutron power plants. The document should be used by organizations designing, manufacturing, constructing and operating nuclear power plants as well as by regulatory bodies

  19. Final safety evaluation report related to the certification of the System 80+ design (Docket No. 52-002). Volume 2, Chapters 15--22 and appendices

    International Nuclear Information System (INIS)

    1994-08-01

    This final safety evaluation report (FSER) documents the technical review of the System 80+ standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the system 80+ design was submitted by Combustion Engineering, Inc., now Asea Brown Boveri-Combustion Engineering (ABB-CE) as an application for design approval and subsequent design certification pursuant to 10 CFR section 52.45. System 80+ is a pressurized water reactor with a rated power of 3914 megawatts thermal (MWt) and a design power of 3992 MWt at which accidents are analyzed. Many features of the System 80+ are similar to those of ABB-CE's System 80 design from which it evolved. Unique features of the System 80+ design include: a large spherical, steel containment; an in-containment refueling water storage tank; a reactor cavity flooding system, hydrogen ignitors and a safety depressurization system for severe accident mitigation; a combustion gas turbine for an alternate ac source; and an advanced digitally based control room. On the basis of its evaluation and independent analyses, the NRC staff concludes that ABB-CE's application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the System 80+ standard design. This document, Volume 2, contains Chapters 15 through 22 and Appendices A through E

  20. Radiation protection aspects in the design of nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2008-01-01

    The IAEA's Statute authorizes the Agency to establish safety standards to protect health and minimize danger to life and property - standards which the IAEA must use in its own operations, and which a State can apply by means of its regulatory provisions for nuclear and radiation safety. A comprehensive body of safety standards under regular review, together with the IAEA's assistance in their application, has become a key element in a global safety regime. In the mid-1990s, a major overhaul of the IAEA's safety standards programme was initiated, with a revised oversight committee structure and a systematic approach to updating the entire corpus of standards. The new standards that have resulted are of a high calibre and reflect best practices in Member States. With the assistance of the Commission on Safety Standards, the IAEA is working to promote the global acceptance and use of its safety standards. Safety standards are only effective, however, if they are properly applied in practice. The IAEA's safety services - which range in scope from engineering safety, operational safety, and radiation, transport and waste safety to regulatory matters and safety culture in organizations - assist Member States in applying the standards and appraise their effectiveness. These safety services enable valuable insights to be shared and continue to urge all Member States to make use of them. Regulating nuclear and radiation safety is a national responsibility, and many Member States have decided to adopt the IAEA's safety standards for use in their national regulations. For the Contracting Parties to the various international safety conventions, IAEA standards provide a consistent, reliable means of ensuring the effective fulfilment of obligations under the conventions. The standards are also applied by designers, manufacturers and operators around the world to enhance nuclear and radiation safety in power generation, medicine, industry, agriculture, research and education

  1. Radiation shielding and safety design

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Ouk; Gil, C. S.; Cho, Y. S.; Kim, D. H.; Kim, H. I.; Kim, J. W.; Lee, C. W.; Kim, K. Y.; Kim, B. H. [KAERI, Daejeon (Korea, Republic of)

    2011-07-15

    A benchmarking for the test facility, evaluations of the prompt radiation fields, evaluation of the induced activities in the facility, and estimation of the radiological impact on the environment were performed in this study. and the radiation safety analysis report for nuclear licensing was written based on this study. In the benchmark calculation, the neutron spectra was measured in the 20 Mev test facility and the measurements were compared with the computational results to verify the calculation system. In the evaluation of the prompt radiation fields, the shielding design for 100 MeV target rooms, evaluations of the leakage doses from the accidents and skyshine analysis were performed. The evaluation of the induced activities were performed for the coolant, inside air, structural materials, soil and ground-water. At last, the radiation safety analysis report was written based on results from these studies

  2. What vehicle features are considered important when buying an automobile? An examination of driver preferences by age and gender.

    Science.gov (United States)

    Vrkljan, Brenda H; Anaby, Dana

    2011-02-01

    Certain vehicle features can help drivers avoid collisions and/or protect occupants in the event of a crash, and therefore, might play an important role when deciding which vehicle to purchase. The objective of this study was to examine the importance attributed to key vehicle features (including safety) that drivers consider when buying a car and its association with age and gender. A sample of 2,002 Canadian drivers aged 18 years and older completed a survey that asked them to rank the importance of eight vehicle features if they were to purchase a vehicle (storage, mileage, safety, price, comfort, performance, design, and reliability). ANOVA tests were performed to: (a) determine if there were differences in the level of importance between features and; (b) examine the effect of age and gender on the importance attributed to these features. Of the features examined, safety and reliability were the most highly rated in terms of importance, whereas design and performance had the lowest rating. Differences in safety and performance across age groups were dependent on gender. This effect was most evident in the youngest and oldest age groups. Safety and reliability were considered the most important features. Age and gender play a significant role in explaining the importance of certain features. Targeted efforts for translating safety-related information to the youngest and oldest consumers should be emphasized due to their high collision, injury, and fatality rates. Copyright © 2011 National Safety Council and Elsevier Ltd. All rights reserved.

  3. Exploring road design factors influencing tram road safety - Melbourne tram driver focus groups.

    Science.gov (United States)

    Naznin, Farhana; Currie, Graham; Logan, David

    2018-01-01

    Melbourne, Australia has the largest tram/streetcar network in the world including the largest mixed traffic tram operating environment. Therefore, Melbourne tram drivers are responsible for controlling one of the heaviest vehicles on road ranging from shared tram lanes to exclusive tram lanes. In addition to different tram lane configurations, tram drivers need to follow different traffic signal phases at intersections including tram priority signals as well as need to serve passengers at various types of closely spaced tram stops. Despite all these challenges, no research has explored tram driver perceptions of the risk factors on different tram route road design configurations. Therefore, the aim of this study is to investigate how tram drivers' safety perceptions alter along various tram route sections, signal settings and stop configurations. A tram driver focus group approach was adopted for this research involving thirty tram drivers (4 female and 26 male drivers). The tram drivers' age ranged from 29 to 63 years, with an average age of 47.6 years (standard deviation of 10.1 years), and their experience of tram driving ranged from 1.17 to 31 years, with an average experience of 12.5 years (standard deviation of 10.2 years). The participating tram drivers perceived that the raised tram tracks and tramways with raised yellow curbing beside tracks are safer lane priority features on the Melbourne tram network compared to full-time, part-time and mixed traffic tram lanes. They regarded 'hook turns' as a safe form of tram signal priority treatment at intersections and platform tram stops as the safest tram stop design for all passengers among all other tram stop designs in Melbourne. Findings of this research could enhance the understanding of crash risk factors for different tram route features and thus can offer effective planning strategies for transit agencies to improve tram road safety. Copyright © 2017 Elsevier Ltd. All rights reserved.

  4. Dhruva: Main design features, operational experience and utilization

    Energy Technology Data Exchange (ETDEWEB)

    Agarwal, S.K. [Reactor Group, BARC, Trombay, Mumbai 400085 (India)]. E-mail: swarajagarwal2000@yahoo.com; Karhadkar, C.G. [Reactor Group, BARC, Trombay, Mumbai 400085 (India); Zope, A.K. [Reactor Group, BARC, Trombay, Mumbai 400085 (India); Singh, Kanchhi [Reactor Group, BARC, Trombay, Mumbai 400085 (India)

    2006-04-15

    Dhruva is a product of technological initiatives taken during mid seventies when a need was felt for another research reactor having a high neutron flux to meet the growing demands of research and development in the frontier areas of science and engineering. In addition production of radioisotopes of high specific activity and the diverse requirements of a broad based user community had to be synthesized into a viable system, which could be engineered within the limited means available in the country. This high neutron flux reactor was designed, constructed and commissioned entirely by Indian scientists and engineers and it reflects the country's resolve to achieve self-reliance in the nuclear reactor technology. Dhruva is a 100 MW (thermal) research reactor with metallic natural uranium as fuel, heavy water as moderator, coolant and reflector, giving a maximum thermal neutron flux of 1.8 x 10{sup 14} n/cm{sup 2}/s. Since its first criticality on 8th August 1985, a number of experimental facilities have been added which have proven to be highly attractive for universities and industrial researchers for their scientific merits in various fields. One of the major utilization areas has been the neutron beam research using several neutron spectrometers, all of which were built in-house. A guide tube facility comprising of two neutron guides and another experimental set-up with a multi-instrument beam line have enabled further enhancement of the utilization of this National Facility by the academic institutions in the country. Production of radioisotopes of high specific activity and in increased quantity has fulfilled growing demands for many applications. The write-up provides an overview of the reactor covering its design; layout, safety features, utilization and operating experience along with description of some of the specific experimental facilities.

  5. Overview of KJRR design features

    International Nuclear Information System (INIS)

    Park, C.; Kim, Y.K.; Lee, B.C.; Ryu, J.S.; Kwon, Y.S.

    2015-01-01

    A new research reactor construction project (hereafter, “KJRR project”) is being conducted in order to secure the supply of key medical and industrial radioisotopes and to develop the core technologies of research reactors such as U-Mo fuel and a reactor bottom mounted control rod drive mechanism. The KJRR project aims to establish a RR with 15 MW and related utilization facilities for RI production and neutron transmutation doping (NTD) services, and the relevant research and development at Kijang-Gun, Busan City in Korea. A preliminary safety analysis report was submitted to the regulatory body for a construction permit in the end of 2014. The KJRR is under a detailed design and is expected to be put into operation in 2019. (author)

  6. Physical design correlates of efficiency and safety in emergency departments: a qualitative examination.

    Science.gov (United States)

    Pati, Debajyoti; Harvey, Thomas E; Pati, Sipra

    2014-01-01

    The objective of this study was to explore and identify physical design correlates of safety and efficiency in emergency department (ED) operations. This study adopted an exploratory, multimeasure approach to (1) examine the interactions between ED operations and physical design at 4 sites and (2) identify domains of physical design decision-making that potentially influence efficiency and safety. Multidisciplinary gaming and semistructured interviews were conducted with stakeholders at each site. Study data suggest that 16 domains of physical design decisions influence safety, efficiency, or both. These include (1) entrance and patient waiting, (2) traffic management, (3) subwaiting or internal waiting areas, (4) triage, (5) examination/treatment area configuration, (6) examination/treatment area centralization versus decentralization, (7) examination/treatment room standardization, (8) adequate space, (9) nurse work space, (10) physician work space, (11) adjacencies and access, (12) equipment room, (13) psych room, (14) staff de-stressing room, (15) hallway width, and (16) results waiting area. Safety and efficiency from a physical environment perspective in ED design are mutually reinforcing concepts--enhancing efficiency bears positive implications for safety. Furthermore, safety and security emerged as correlated concepts, with security issues bearing implications for safety, thereby suggesting important associations between safety, security, and efficiency.

  7. Graphical symbols -- Safety colours and safety signs -- Part 1: Design principles for safety signs in workplaces and public areas

    CERN Document Server

    International Organization for Standardization. Geneva

    2002-01-01

    This International Standard establishes the safety identification colours and design principles for safety signs to be used in workplaces and in public areas for the purpose of accident prevention, fire protection, health hazard information and emergency evacuation. It also establishes the basic principles to be applied when developing standards containing safety signs. This part of ISO 3864 is applicable to workplaces and all locations and all sectors where safety-related questions may be posed. However, it is not applicable to the signalling used for guiding rail, road, river, maritime and air traffic and, generally speaking, to those sectors subject to a regulation which may differ.

  8. Children's choice: Color associations in children's safety sign design.

    Science.gov (United States)

    Siu, Kin Wai Michael; Lam, Mei Seung; Wong, Yi Lin

    2017-03-01

    Color has been more identified as a key consideration in ergonomics. Color conveys messages and is an important element in safety signs, as it provides extra information to users. However, very limited recent research has focused on children and their color association in the context of safety signs. This study thus examined how children use colors in drawing different safety signs and how they associate colors with different concepts and objects that appear in safety signs. Drawing was used to extract children's use of color and the associations they made between signs and colors. The child participants were given 12 referents of different safety signs and were asked to design and draw the signs using different colored felt-tip pens. They were also asked to give reasons for their choices of colors. Significant associations were found between red and 'don't', orange and 'hands', and blue and 'water'. The child participants were only able to attribute the reasons for the use of yellow, green, blue and black through concrete identification and concrete association, and red through abstract association. The children's use of color quite differs from that shown in the ISO registered signs. There is a need to consider the use of colors carefully when designing signs specifically for children. Sign designers should take children's color associations in consideration and be aware if there are any misunderstandings. Copyright © 2016 Elsevier Ltd. All rights reserved.

  9. Laser safety in design of near-infrared scanning LIDARs

    Science.gov (United States)

    Zhu, X.; Elgin, D.

    2015-05-01

    3D LIDARs (Light Detection and Ranging) with 1.5μm nanosecond pulse lasers have been increasingly used in different applications. The main reason for their popularity is that these LIDARs have high performance while at the same time can be made eye-safe. Because the laser hazard effect on eyes or skin at this wavelength region (industrial mining applications. We have incorporated the laser safety requirements in the LIDAR design and conducted laser safety analysis for different operational scenarios. While 1.5μm is normally said to be the eye-safe wavelength, in reality a high performance 3D LIDAR needs high pulse energy, small beam size and high pulse repetition frequency (PRF) to achieve long range, high resolution and high density images. The resulting radiant exposure of its stationary beam could be many times higher than the limit for a Class 1 laser device. Without carefully choosing laser and scanning parameters, including field-of-view, scan speed and pattern, a scanning LIDAR can't be eye- or skin-safe based only on its wavelength. This paper discusses the laser safety considerations in the design of eye-safe scanning LIDARs, including laser pulse energy, PRF, beam size and scanning parameters in two basic designs of scanning mechanisms, i.e. galvanometer based scanner and Risley prism based scanner. The laser safety is discussed in terms of device classification, nominal ocular hazard distance (NOHD) and safety glasses optical density (OD).

  10. International cooperation in the safety and environmental assessment for the ITER engineering design activities

    International Nuclear Information System (INIS)

    Gordon, C.; Baker, D.J.; Bartels, H-W.

    1998-01-01

    The ITER Project includes design and assessment activities to ensure the safety and environmental attractiveness of ITER and demonstrate that it can be sited in any of the sponsoring Parties with a minimum of site-specific redesign. This paper highlights some of the efforts to develop an international consensus approach for ITER safety design and assessment, including: development of general safety and environmental design criteria; development of quantitative dose-release assessment criteria; development of a radiation protection program; waste characterization; and development of safety analysis guidelines. The high level of interaction, cooperation and collaboration between the Joint Central Team and the Home Teams, and between the safety team and designers, and the spirit of consensus that has guided them have resulted in a safe design for ITER and a safety design and assessment that can meet the needs of the potential host countries. (author)

  11. External Events Excluding Earthquakes in the Design of Nuclear Power Plants. Safety Guide

    International Nuclear Information System (INIS)

    2008-01-01

    This Safety Guide provides recommendations and guidance on design for the protection of nuclear power plants from the effects of external events (excluding earthquakes), i.e. events that originate either off the site or within the boundaries of the site but from sources that are not directly involved in the operational states of the nuclear power plant units. In addition, it provides recommendations on engineering related matters in order to comply with the safety objectives and requirements established in the IAEA Safety Requirements publication, Safety of Nuclear Power Plants: Design. It is also applicable to the design and safety assessment of items important to the safety of land based stationary nuclear power plants with water cooled reactors. Contents: 1. Introduction; 2. Application of safety criteria to the design; 3. Design basis for external events; 4. Aircraft crash; 5. External fire; 6. Explosions; 7. Asphyxiant and toxic gases; 8. Corrosive and radioactive gases and liquids; 9. Electromagnetic interference; 10. Floods; 11. Extreme winds; 12. Extreme meteorological conditions; 13. Biological phenomena; 14. Volcanism; 15. Collisions of floating bodies with water intakes and UHS components; Annex I: Aircraft crashes; Annex II: Detonation and deflagration; Annex III: Toxicity limits.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

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

    International Nuclear Information System (INIS)

    Park, Jaekwan; Suh, Yongsuk

    2014-01-01

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

  14. Introduction of Autonomous Vehicles: Roundabouts Design and Safety Performance Evaluation

    Directory of Open Access Journals (Sweden)

    Aleksandra Deluka Tibljaš

    2018-04-01

    Full Text Available Driving experiences provided by the introduction of new vehicle technologies are directly impacting the criteria for road network design. New criteria should be taken into consideration by designers, researchers and car owners in order to assure traffic safety in changed conditions that will appear with, for example, introduction of Autonomous Vehicles (AVs in everyday traffic. In this paper, roundabout safety level is analysed on the originally developed microsimulation model in circumstances where different numbers of AVs vehicles are mixed with Conventional Vehicles (CVs. Field data about speed and traffic volumes from existing roundabouts in Croatia were used for development of the model. The simulations done with the Surrogate Safety Assessment Model (SSAM give some relevant highlights on how the introduction of AVs could change both operational and safety parameters at roundabouts. To further explore the effects on safety of roundabouts with the introduction of different shares of AVs, hypothetical safety treatments could be tested to explore whether their effects may change, leading to the estimation of a new set of Crash Modification Factors.

  15. Examining the design features of a communication-rich, problem-centred mathematics professional development

    Science.gov (United States)

    de Araujo, Zandra; Orrill, Chandra Hawley; Jacobson, Erik

    2018-04-01

    While there is considerable scholarship describing principles for effective professional development, there have been few attempts to examine these principles in practice. In this paper, we identify and examine the particular design features of a mathematics professional development experience provided for middle grades teachers over 14 weeks. The professional development was grounded in a set of mathematical tasks that each had one right answer, but multiple solution paths. The facilitator engaged participants in problem solving and encouraged participants to work collaboratively to explore different solution paths. Through analysis of this collaborative learning environment, we identified five design features for supporting teacher learning of important mathematics and pedagogy in a problem-solving setting. We discuss these design features in depth and illustrate them by presenting an elaborated example from the professional development. This study extends the existing guidance for the design of professional development by examining and operationalizing the relationships among research-based features of effective professional development and the enacted features of a particular design.

  16. Safety and design impact of hurricane Andrew

    International Nuclear Information System (INIS)

    Guey, Ching N.

    2004-01-01

    Turkey Point completed the IPE in June of 1991. Hurricane Andrew landed at Turkey Point on August 24, 1992. Although the safety related systems, components and structures were not damaged by the Hurricane Andrew, certain nonsafety related components and the neighboring fossil plant sustained noticeable damage. Among the major components that were nonsafety related but would affect the PRA of the plant included the service water pumps and the high tower. This paper discusses the safety and design impact of Hurricane Andrew on Turkey Point Nuclear Power Plant. The risk of hurricanes on the interim and evolving plant configurations are briefly described. The risk of the plant from internal events as a result of damage incurred during Hurricane Andrew are discussed. The design change as the result of Hurricane Andrew and its impact on the PRA are presented. (author)

  17. New features in the design code TLIE

    International Nuclear Information System (INIS)

    van Zeijts, J.

    1993-01-01

    We present features recently installed in the arbitrary-order accelerator design code TLIE. The code uses the MAD input language, and implements programmable extensions modeled after the C language that make it a powerful tool in a wide range of applications: from basic beamline design to high precision-high order design and even control room applications. The basic quantities important in accelerator design are easily accessible from inside the control language. Entities like parameters in elements (strength, current), transfer maps (either in Taylor series or in Lie algebraic form), lines, and beams (either as sets of particles or as distributions) are among the type of variables available. These variables can be set, used as arguments in subroutines, or just typed out. The code is easily extensible with new datatypes

  18. Small Column Ion Exchange Design and Safety Strategy

    International Nuclear Information System (INIS)

    Huff, T.; Rios-Armstrong, M.; Edwards, R.; Herman, D.

    2011-01-01

    Small Column Ion Exchange (SCIX) is a transformational technology originally developed by the Department of Energy (DOE) Environmental Management (EM-30) office and is now being deployed at the Savannah River Site (SRS) to significantly increase overall salt processing capacity and accelerate the Liquid Waste System life-cycle. The process combines strontium and actinide removal using Monosodium Titanate (MST), Rotary Microfiltration, and cesium removal using Crystalline Silicotitanate (CST, specifically UOP IONSIV(reg s ign)IE-911 ion exchanger) to create a low level waste stream to be disposed in grout and a high level waste stream to be vitrified. The process also includes preparation of the streams for disposal, e.g., grinding of the loaded CST material. These waste processing components are technically mature and flowsheet integration studies are being performed including glass formulations studies, application specific thermal modeling, and mixing studies. The deployment program includes design and fabrication of the Rotary Microfilter (RMF) assembly, ion-exchange columns (IXCs), and grinder module, utilizing an integrated system safety design approach. The design concept is to install the process inside an existing waste tank, Tank 41H. The process consists of a feed pump with a set of four RMFs, two IXCs, a media grinder, three Submersible Mixer Pumps (SMPs), and all supporting infrastructure including media receipt and preparation facilities. The design addresses MST mixing to achieve the required strontium and actinide removal and to prevent future retrieval problems. CST achieves very high cesium loadings (up to 1,100 curies per gallon (Ci/gal) bed volume). The design addresses the hazards associated with this material including heat management (in column and in-tank), as detailed in the thermal modeling. The CST must be size reduced for compatibility with downstream processes. The design addresses material transport into and out of the grinder and

  19. Development of safety analysis technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sim, Suk K.; Song, J. H.; Chung, Y. J. and others

    1999-03-01

    Inherent safety features and safety system characteristics of the SMART integral reactor are investigated in this study. Performance and safety of the SMART conceptual design have been evaluated and confirmed through the performance and safety analyses using safety analysis system codes as well as a preliminary performance and safety analysis methodology. SMART design base events and their acceptance criteria are identified to develop a preliminary PIRT for the SMART integral reactor. Using the preliminary PIRT, a set of experimental program for the thermal hydraulic separate effect tests and the integral effect tests was developed for the thermal hydraulic model development and the system code validation. Safety characteristics as well as the safety issues of the integral reactor has been identified during the study, which will be used to resolve the safety issues and guide the regulatory criteria for the integral reactor. The results of the performance and safety analyses performed during the study were used to feedback for the SMART conceptual design. The performance and safety analysis code systems as well as the preliminary safety analysis methodology developed in this study will be validated as the SMART design evolves. The performance and safety analysis technology developed during the study will be utilized for the SMART basic design development. (author)

  20. Joint Feature Extraction and Classifier Design for ECG-Based Biometric Recognition.

    Science.gov (United States)

    Gutta, Sandeep; Cheng, Qi

    2016-03-01

    Traditional biometric recognition systems often utilize physiological traits such as fingerprint, face, iris, etc. Recent years have seen a growing interest in electrocardiogram (ECG)-based biometric recognition techniques, especially in the field of clinical medicine. In existing ECG-based biometric recognition methods, feature extraction and classifier design are usually performed separately. In this paper, a multitask learning approach is proposed, in which feature extraction and classifier design are carried out simultaneously. Weights are assigned to the features within the kernel of each task. We decompose the matrix consisting of all the feature weights into sparse and low-rank components. The sparse component determines the features that are relevant to identify each individual, and the low-rank component determines the common feature subspace that is relevant to identify all the subjects. A fast optimization algorithm is developed, which requires only the first-order information. The performance of the proposed approach is demonstrated through experiments using the MIT-BIH Normal Sinus Rhythm database.

  1. The Impact Of Website Design Features On Behavioral Intentions

    Directory of Open Access Journals (Sweden)

    Chun-Chin Chiu

    2015-08-01

    Full Text Available The design of a website interface plays an important role in online purchasing and customers are more likely to visit and buy from better-designed websites. However previous studies have not provided consistent information about the features a website should provide. Based on Hausman and Siekpes 2009 comprehensive model this study aims to empirically verify whether the model can be applied in e-service markets to predict and explain website users behavioral intentions trade intentions and revisit intentions. Based on the data from a survey of 303 Internet users the results indicate that computer factors and human factors the key website design features are significantly related to website users experiences perceived usefulness perceived entertainment value and perceived informativeness in turn significantly affect the intermediary outcomes of attitude toward the site and ultimately influence users behavioral intentions.

  2. Review of SFR Design Safety using Preliminary Regulatory PSA Model

    International Nuclear Information System (INIS)

    Na, Hyun Ju; Lee, Yong Suk; Shin, Andong; Suh, Nam Duk

    2013-01-01

    The major objective of this research is to develop a risk model for regulatory verification of the SFR design, and thereby, make sure that the SFR design is adequate from a risk perspective. In this paper, the development result of preliminary regulatory PSA model of SFR is discussed. In this paper, development and quantification result of preliminary regulatory PSA model of SFR is discussed. It was confirmed that the importance PDRC and ADRC dampers is significant as stated in the result of KAERI PSA model. However, the importance can be changed significantly depending on assumption of CCCG and CCF factor of PDRC and ADRC dampers. SFR (sodium-cooled fast reactor) which is Gen-IV nuclear energy system, is designed to accord with the concept of stability, sustainability and proliferation resistance. KALIMER-600, which is under development in Korea, includes passive safety systems (e. g. passive reactor shutdown, passive residual heat removal, and etc.) as well as active safety systems. Risk analysis from a regulatory perspective is needed to support the regulatory body in its safety and licensing review for SFR (KALIMER-600). Safety issues should be identified in the early design phase in order to prevent the unexpected cost increase and delay of the SFR licensing schedule that may be caused otherwise

  3. Development of Safety Analysis Technology for Integral Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sim, S. K. [Korea Atomic Energy Research Institute, Taejeon (Korea); Seul, K. W.; Kim, W. S.; Kim, W. K.; Yun, Y. G.; Ahn, H. J.; Lee, J. S.; Sin, A. D. [Korea Institute of Nuclear Safety, Taejeon (Korea)

    2000-03-01

    The Nuclear Desalination Plant(NDP) is being developed to produce electricity and fresh water, and is expected to locate near population zone. In the aspect of safety, it is required to protect the public and environment from the possible releases of fission products and to prevent the fresh water from the contamination of radioactivity. Thus, in a present study, the safety characteristics of the integral reactor adopting passive and inherent safety features significantly different from existing nuclear power plants were investigated based on the design of foreign and domestic integral reactors. Also, safety requirements applicable to the NDP were analyzed based on the regulatory requirements for current and advanced reactor designs, and use requirements for small-medium size reactors. Based on these analyses, some safety concerns to be considered in the design stage have been identified. They includes the use of proven technology for new safety systems, the systematic classification and selection of design basis accidents, and the safety assurance of desalination-related systems. These efforts to identify and resolve the safety concerns in the design stage will provide the early confidence of SMART safety to designers, and the technical basis to evaluate the safety to reviewers in the future. 8 refs., 20 figs., 4 tabs. (Author)

  4. Safety in the ARIES Tokamak Design Study

    International Nuclear Information System (INIS)

    Herring, J.S.; Wong, C.P.-C.; Cheng, E.T.; Grotz, S.

    1989-01-01

    Safety is one of the primary goals of the ARIES Tokamak Design Study. Public safety goals are the achievement passive safety which is demonstrable in tests that could precede operation and the assurance that releases from accidents be passively limited such that no evacuation plan in necessary. Strategies for safety of the plant investment are factory fabrication, short construction times and a design such that no off-normal operational transient results in damage which could not be repaired in routine maintenance. ARIES-I, the first of three 'visions' of potential tokamak reactors, will use He at 5 MPa as a blanket coolant and SiC/composite ceramic for the first wall and blanket materials. Both the coolant and the structural material were chosen for their low activation, both in the short term after accidents and for long term waste management. The breeder, Li 4 SiO 4 , was also chosen for low activation. Contemporary plasma physics and aggressive technology are used in ARIES-I, which results in very high toroidal fields (24 T maximum at the coil). The stored TF energy will be about 130 GJ. A central concern is the safe discharge of this stored energy under electrical fault conditions and prevention of a failure in the magnet set from propagating into systems containing radioactive inventories. The TF coil system consists of 16 coils, each containing two separate windings powered by two independent power supplies. Arcs and shorts between the two power supply systems and across individual windings have been modeled. In addition, delay or failure in circuit breaker opening has been modeled. The safety impacts of LOCA, LOFA and disruptive events have also been evaluated. 8 refs., 4 figs., 7 tabs

  5. Design features of the radioactive Liquid-Fed Ceramic Melter system

    International Nuclear Information System (INIS)

    Holton, L.K. Jr.

    1985-06-01

    During 1983, the Pacific Northwest Laboratory (PNL), at the request of the Department of Energy (DOE), undertook a program with the principal objective of testing the Liquid-Fed Ceramic Melter (LFCM) process in actual radioactive operations. This activity, termed the Radioactive LFCM (RLFCM) Operations is being conducted in existing shielded hot-cell facilities in B-Cell of the 324 Building, 300 Area, located at Hanford, Washington. This report summarizes the design features of the RLFCM system. These features include: a waste preparation and feed system which uses pulse-agitated waste preparation tanks for waste slurry agitation and an air displacement slurry pump for transferring waste slurries to the LFCM; a waste vitrification system (LFCM) - the design features, design approach, and reasoning for the design of the LFCM are described; a canister-handling turntable for positioning canisters underneath the RLFCM discharge port; a gamma source positioning and detection system for monitoring the glass fill level of the product canisters; and a primary off-gas treatment system for removing the majority of the radionuclide contamination from the RLFCM off gas. 8 refs., 48 figs., 6 tabs

  6. Final safety evaluation report related to the certification of the System 80{sup +} design (Docket No. 52-002). Volume 2, Chapters 15--22 and appendices

    Energy Technology Data Exchange (ETDEWEB)

    1994-08-01

    This final safety evaluation report (FSER) documents the technical review of the System 80+ standard design by the US Nuclear Regulatory Commission (NRC) staff. The application for the system 80+ design was submitted by Combustion Engineering, Inc., now Asea Brown Boveri-Combustion Engineering (ABB-CE) as an application for design approval and subsequent design certification pursuant to 10 CFR {section} 52.45. System 80+ is a pressurized water reactor with a rated power of 3914 megawatts thermal (MWt) and a design power of 3992 MWt at which accidents are analyzed. Many features of the System 80+ are similar to those of ABB-CE`s System 80 design from which it evolved. Unique features of the System 80+ design include: a large spherical, steel containment; an in-containment refueling water storage tank; a reactor cavity flooding system, hydrogen ignitors and a safety depressurization system for severe accident mitigation; a combustion gas turbine for an alternate ac source; and an advanced digitally based control room. On the basis of its evaluation and independent analyses, the NRC staff concludes that ABB-CE`s application for design certification meets the requirements of Subpart B of 10 CFR Part 52 that are applicable and technically relevant to the System 80+ standard design. This document, Volume 2, contains Chapters 15 through 22 and Appendices A through E.

  7. Software Design Improvements. Part 2; Software Quality and the Design and Inspection Process

    Science.gov (United States)

    Lalli, Vincent R.; Packard, Michael H.; Ziemianski, Tom

    1997-01-01

    The application of assurance engineering techniques improves the duration of failure-free performance of software. The totality of features and characteristics of a software product are what determine its ability to satisfy customer needs. Software in safety-critical systems is very important to NASA. We follow the System Safety Working Groups definition for system safety software as: 'The optimization of system safety in the design, development, use and maintenance of software and its integration with safety-critical systems in an operational environment. 'If it is not safe, say so' has become our motto. This paper goes over methods that have been used by NASA to make software design improvements by focusing on software quality and the design and inspection process.

  8. Generalized railway tank car safety design optimization for hazardous materials transport: Addressing the trade-off between transportation efficiency and safety

    International Nuclear Information System (INIS)

    Saat, Mohd Rapik; Barkan, Christopher P.L.

    2011-01-01

    North America railways offer safe and generally the most economical means of long distance transport of hazardous materials. Nevertheless, in the event of a train accident releases of these materials can pose substantial risk to human health, property or the environment. The majority of railway shipments of hazardous materials are in tank cars. Improving the safety design of these cars to make them more robust in accidents generally increases their weight thereby reducing their capacity and consequent transportation efficiency. This paper presents a generalized tank car safety design optimization model that addresses this tradeoff. The optimization model enables evaluation of each element of tank car safety design, independently and in combination with one another. We present the optimization model by identifying a set of Pareto-optimal solutions for a baseline tank car design in a bicriteria decision problem. This model provides a quantitative framework for a rational decision-making process involving tank car safety design enhancements to reduce the risk of transporting hazardous materials.

  9. Role of Passive Safety Features in Prevention And Mitigation of Severe Plant Conditions in Indian Advanced Heavy Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jain, Vikas; Nayak, A.; Dhiman, M.; Kulkarni, P. P.; Vijayan, P. K.; Vaze, K. K. [Bhabha Atomic Research Centre, Mumbai (India)

    2013-10-15

    Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor 'AHWR' is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI), Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

  10. ROLE OF PASSIVE SAFETY FEATURES IN PREVENTION AND MITIGATION OF SEVERE PLANT CONDITIONS IN INDIAN ADVANCED HEAVY WATER REACTOR

    Directory of Open Access Journals (Sweden)

    VIKAS JAIN

    2013-10-01

    Full Text Available Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor ‘AHWR’ is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI, Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

  11. Improvement of operation efficiency for WWER-440 and WWER-1000 for TRIGON fuel assembly design features

    Energy Technology Data Exchange (ETDEWEB)

    Silberstein, A [European WWER Fuels GmbH, Lyon (France)

    1994-12-31

    TRIGON 440 and TRIGON 1000 fuel assemblies and their assembly matching counterparts are described. Their role in increasing the efficiency of WWER reactors is stressed. Special attention is paid to their design features as well as calibrated means of predicting behaviour under irradiation from light water reactor core operation. They reduce the fuel cycle cost as a result of the reduced need for natural uranium which have to be enriched and of the smaller number of fuel assemblies which have to be fabricated, stored or reprocessed. The improved control assemblies bring comfort to the plant operator due to intrinsic progress in safety with respect to accidental situation, trouble-free behaviour and long time utilization in the reactor. 14 figs.

  12. FEATURES ROAD SAFETY AUDIT

    Directory of Open Access Journals (Sweden)

    L. Abramova

    2015-07-01

    Full Text Available Development of the road network, increasing motorization of the population significantly increase the risk of accidents. Experts in the field of traffic are developing methods to reduce the probability of accidents. The ways of solving the problems of road safety audit at various stages of the «life» of roads are considered.

  13. Integral type small PWR with stand-alone safety

    International Nuclear Information System (INIS)

    Makihara, Yoshiaki

    2001-01-01

    A feasibility study is achieved on an integral type small PWR with stand-alone safety. It is designed to have the following features. (1) The coolant does not leak out at any accidental condition. (2) The fuel failure does never occur while it is supposed on the large scale PWR at the design base accident. (3) At any accidental condition the safety is secured without any support from the outside (stand-alone safety secure). (4) It has self-regulating characteristics and easy controllability. The above features can be satisfied by integrate the steam generator and CRDM in the reactor vessel while the pipe line break has to be considered on the conventional PWR. Several counter measures are planned to satisfy the above features. The economy feature is also attained by several simplifications such as (1) elimination of main coolant piping and pressurizer by the integration of primary cooling system and self-pressurizing, (2) elimination of RCP by application of natural circulating system, (3) elimination of ECCS and accumulator by application of static safety system, (4) large scale volume reduction of the container vessel by application of integrated primary cooling system, (5) elimination of boric acid treatment by deletion of chemical shim. The long operation period such as 10 years can be attained by the application of Gd fuel in one batch refueling. The construction period can be shortened by the standardizing the design and the introduction of modular component system. Furthermore the applicability of the reduced modulation core is also considered. (K. Tsuchihashi)

  14. Criticality safety and facility design considerations

    International Nuclear Information System (INIS)

    Waltz, W.R.

    1991-06-01

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

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

    International Nuclear Information System (INIS)

    Ingersoll, Daniel T.

    2011-01-01

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

  16. Advanced nuclear reactor safety design technology research in NPIC

    International Nuclear Information System (INIS)

    Yu, H.

    2014-01-01

    After the Fukushima accident happen, Nuclear Power Plants (NPPs) construction has been suspended in China for a time. Now the new regulatory rule has been proposed that the most advanced safety standard must be adopted for the new NPPs and practical elimination of large fission product release by design during the next five plans period. So the advanced reactor research is developing in China. NPIC is engaging on the ACP1000 and ACP100 (Small Module Reactor) design. The main design character will be introduced in this paper. The Passive Combined with Active (PCWA) design was adopted during the ACP1000 design to reduce the core damage frequency (CDF); the Cavity Injection System (CIS) is design to mitigation the consequence of the severe accident. Advance passive safety system was designed to ensure the long term residual heat removal during the Small Module Reactor (SMR). The SMR will be utilized to be the floating reactors, district heating reactor and so on. Besides, the Science and Technology on Reactor System Design Technology Laboratory (LRSDT) also engaged on the fundamental thermal-hydraulic characteristic research in support of the system validation. (author)

  17. Integrated Safety in ''SARAF'

    International Nuclear Information System (INIS)

    Dickstein, P.; Grof, Y.; Machlev, M.; Pernick, A.

    2004-01-01

    As of the very early stages of the accelerator project at the Soreq Nuclear Research Center ''SARAF'' a safety group was established which has been an inseparable participant in the planning and design of the new facility. The safety group comprises of teams responsible for the shielding, radiation protection and general industrial safety aspects of ''SARAF''. The safety group prepared and documented the safety envelope for the accelerator, dealing with the safety requirements and guidelines for the first, pre-operational, stages of the project. The safety envelope, though based upon generic principles, took into account the accelerator features and the expected modes of operation. The safety envelope was prepared in a hierarchical structure, containing Basic Principles, Basic Guidelines, General Principles for Safety Implementation, Safety Requirements and Safety Underlining Issues. The above safety envelope applies to the entire facility, which entails the accelerator itself and the experimental areas and associated plant and equipment utilizing and supporting the production of the accelerated particle beams

  18. Analysis of French (Paluel) pressurized water reactor design differences compared to current US PWR designs

    International Nuclear Information System (INIS)

    1986-05-01

    To understand better the regulatory approaches to reactor safety in foreign countries, the staff of the Nuclear Regulatory Commisssion has reviewed design information on the Paluel nuclear power plant, one of the current standard 1300-MWe plant operating in France. This report provides the staff's evaluation of major design differences between this standardized French plant and current US pressurized water reactor plants, as well as insights concerning French regulatory practices. The staff identified approximately 25 design differences, and an analysis of the safety significance of each of these design features is presented, along with an assessment comparing the relative safety benefit of each

  19. A series of student design projects for improving and modernizing safety helmets

    NARCIS (Netherlands)

    Beurden, van K.M.M. (Karin); Boer, de J. (Johannes); Stilma, M. (Margot); Teeuw, W.B. (Wouter)

    2014-01-01

    The Saxion Research Centre for Design and Technology employs many students during research projects. This paper discusses a series of student design projects on safety helmets in the Safety@Work project. At construction sites workers are required to wear personal protective equipment during their

  20. Verification of Overall Safety Factors In Deterministic Design Of Model Tested Breakwaters

    DEFF Research Database (Denmark)

    Burcharth, H. F.

    2001-01-01

    The paper deals with concepts of safety implementation in design. An overall safety factor concept is evaluated on the basis of a reliability analysis of a model tested rubble mound breakwater with monolithic super structure. Also discussed are design load identification and failure mode limit...

  1. Design Safety Considerations for Water Cooled Small Modular Reactors Incorporating Lessons Learned from the Fukushima Daiichi Accident

    International Nuclear Information System (INIS)

    2016-03-01

    The global future deployment of advanced nuclear reactors for electricity generation depends primarily on the ability of nuclear industries, utilities and regulatory authorities to further enhance their reliability and economic competitiveness while satisfying stringent safety requirements. The IAEA has a project to help coordinate Member States efforts in the development and deployment of small and medium sized or small modular reactor (SMR) technology. This project aims simultaneously to facilitate SMR technology developers and potential SMR uses, particularly States embarking on a nuclear power programme, in identifying key enabling technologies and enhancing capacity building by resolving issues relevant to deployment, including nuclear reactor safety. The objective of this publication is to explore common practices for Member States, which will be an essential resource for future development and deployment of SMR technology. The accident at the Fukushima Daiichi nuclear power plant was caused by an unprecedented combination of natural events: a strong earthquake, beyond the design basis, followed by a series of tsunamis of heights exceeding the design basis tsunami considered in the flood analysis for the site. Consequently, all the operating nuclear power plants and advanced reactors under development, including SMRs, have been incorporating lessons learned from the accident to assure and enhance the performance of the engineered safety features in coping with such external events

  2. MAPLE research reactor safety uncertainty assessment methodology

    International Nuclear Information System (INIS)

    Sills, H.E.; Duffey, R.B.; Andres, T.H.

    1999-01-01

    The MAPLE (multipurpose Applied Physics Lattice Experiment) reactor is a low pressure, low temperature, open-tank-in pool type research reactor that operates at a power level of 5 to 35 MW. MAPLE is designed for ease of operation, maintenance, and to meet today's most demanding requirements for safety and licensing. The emphasis is on the use of passive safety systems and environmentally qualified components. Key safety features include two independent and diverse shutdown systems, two parallel and independent cooling loops, fail safe operation, and a building design that incorporates the concepts of primary containment supported by secondary confinement

  3. Design features of an automated entry control system

    International Nuclear Information System (INIS)

    Reynolds, D.A.

    1978-01-01

    Features of an entry control system designed to automatically control access to nuclear facilities is described. Control independent of variable human factors is stressed, but security force action is required for assessment and response as a result of an alarm. A design based on a distributed processing capability is utilized. Flexibility and generality are emphasized in an effort to maximize applicability to the entry-control problem faced by nuclear facilities upgrading security as a result of the Safeguards Program

  4. Advanced analysis and design for fire safety of steel structures

    CERN Document Server

    Li, Guoqiang

    2013-01-01

    Advanced Analysis and Design for Fire Safety of Steel Structures systematically presents the latest findings on behaviours of steel structural components in a fire, such as the catenary actions of restrained steel beams, the design methods for restrained steel columns, and the membrane actions of concrete floor slabs with steel decks. Using a systematic description of structural fire safety engineering principles, the authors illustrate the important difference between behaviours of an isolated structural element and the restrained component in a complete structure under fire conditions. The book will be an essential resource for structural engineers who wish to improve their understanding of steel buildings exposed to fires. It is also an ideal textbook for introductory courses in fire safety for master’s degree programs in structural engineering, and is excellent reading material for final-year undergraduate students in civil engineering and fire safety engineering. Furthermore, it successfully bridges th...

  5. Improvement of inherent safety features in CSR (Coupled Spectrum Reactor) for treating MA

    International Nuclear Information System (INIS)

    Aziz, F.; Kitamoto, Asashi.

    1996-01-01

    Burning and/or transmutation (B/T) of MA is proposed here using a CSR (Coupled Spectrum Reactor) concept. CSR was based on a modified conventional 1150 MWe-PWR system, and consisted of two core regions for thermal and fast neutrons, respectively. The B/T fuel used was supposed such that MA discharged from 1 GWe-LWR were mixed homogeneously in LWR fuel. The geometry of B/T fuel in the outer region was left the same with that of PWR, while in the inner region the B/T fuel was arranged in a tight-lattice geometry that allowed a higher fuel to coolant volume ratio, (V m /V f ). In order to improve its inherent safety features, several cases of CSR were studied and compared, each case used different fuel type in the inner region. The result of the calculations showed that safety features can be improved by using composite fuel of ( 235 U-Pu- 238 U) in the inner region. The equilibrium of main isotopes in CSR can be achieved after about 5 recycle stages. This study also showed that the CSR can burn and transmute MA up to 808 kg/stage in a single reactor operated with a reactivity swing of 2.8 % Δk/kk'. (author)

  6. A new safety approach in the design of fast reactors

    International Nuclear Information System (INIS)

    Neuhold, R.J.; Marchaterre, J.F.; Waltar, A.E.

    1987-01-01

    A new approach to achieving fast reactor safety goals is becoming really apparent in the US Fast Reactor Program. Whereas the ''defense is best'' philosophy still prevails, there has been a tangible shift toward emphasizing passive mechanisms to protect the reactor and provide public safety---rather than relying on add-on active, engineered safety systems. This paper reviews the technical basis for this new safety approach and provides discussion on its implementation in current US liquid metal-cooled reactor designs. 4 refs., 4 figs

  7. Safety considerations and countermeasures against fire and explosion at an HTGR-hydrogen production system. Proposal of safety design concept

    International Nuclear Information System (INIS)

    Nishihara, T.; Hada, K.; Shibata, T.; Shiozawa, S.

    1996-01-01

    Establishment of safety design concept and countermeasures against fire and explosion accidents is among key safety-related issues in an HTGR-hydrogen production system. We propose the different safety design concepts depending upon the origin of fire and explosion which may happen in the HTGR-hydrogen production plant. Against fire and explosion originated outside the reactor building (R/B), namely in the area of hydrogen production plant, the safety design concept is primarily to take a safe distance for preventing the damage on safety-related items or a proof wall if necessary. Because the hydrogen production plant is designed in the same safety level as a conventional chemical plant. The safe distance is proposed to limit an incident overpressure to 10 kPa so as not to suffer any damage on the items and to limit a wall-averaged temperature of concrete structures of the R/B to 175degC according to the current regulation. On the other hand, against a potential possibility of explosion originated inside the R/B, the safety design concept is to minimize the possibility of explosion low enough to assume no occurrence inside the R/B. That is, the measure is to exclude a simultaneous failure of a secondary helium piping and an endothermic chemical reactor. Furthermore, in severe accident condition in which the explosion may be postulated a priori, an incidental overpressure of explosion inside the reactor containment vessel (C/V) should be limited so as not to fail the C/V through restricting the amount of combustible gas ingress into the C/V by means of a combination of C/V isolation valve installed in the helium piping and emergency shut off valve in the process feed gas line. (author)

  8. SYSTEMS SAFETY ANALYSIS FOR FIRE EVENTS ASSOCIATED WITH THE ECRB CROSS DRIFT

    International Nuclear Information System (INIS)

    R. J. Garrett

    2001-01-01

    The purpose of this analysis is to systematically identify and evaluate fire hazards related to the Yucca Mountain Site Characterization Project (YMP) Enhanced Characterization of the Repository Block (ECRB) East-West Cross Drift (commonly referred to as the ECRB Cross-Drift). This analysis builds upon prior Exploratory Studies Facility (ESF) System Safety Analyses and incorporates Topopah Springs (TS) Main Drift fire scenarios and ECRB Cross-Drift fire scenarios. Accident scenarios involving the fires in the Main Drift and the ECRB Cross-Drift were previously evaluated in ''Topopah Springs Main Drift System Safety Analysis'' (CRWMS M and O 1995) and the ''Yucca Mountain Site Characterization Project East-West Drift System Safety Analysis'' (CRWMS M and O 1998). In addition to listing required mitigation/control features, this analysis identifies the potential need for procedures and training as part of defense-in-depth mitigation/control features. The inclusion of this information in the System Safety Analysis (SSA) is intended to assist the organization(s) (e.g., Construction, Environmental Safety and Health, Design) responsible for these aspects of the ECRB Cross-Drift in developing mitigation/control features for fire events, including Emergency Refuge Station(s). This SSA was prepared, in part, in response to Condition/Issue Identification and Reporting/Resolution System (CIRS) item 1966. The SSA is an integral part of the systems engineering process, whereby safety is considered during planning, design, testing, and construction. A largely qualitative approach is used which incorporates operating experiences and recommendations from vendors, the constructor and the operating contractor. The risk assessment in this analysis characterizes the scenarios associated with fires in terms of relative risk and includes recommendations for mitigating all identified hazards. The priority for recommending and implementing mitigation control features is: (1) Incorporate

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  10. A proposal for safety design philosophy of HTGR for coupling hydrogen production plant

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Ohashi, Hirofumi; Tazawa, Yujiro; Imai, Yoshiyuki; Nakagawa, Shigeaki; Tachibana, Yukio; Kunitomi, Kazuhiko

    2013-06-01

    Japan Atomic Energy Agency (JAEA) has been conducting research and development for hydrogen production utilizing heat from High Temperature Gas-cooled Reactors (HTGRs). Towards the realization of nuclear hydrogen production, coupled hydrogen production plants should not be treated as an extension of a nuclear plant in order to open the door for the entry of non-nuclear industries as well as assuring reactor safety against postulated abnormal events initiated in the hydrogen production plants. Since hydrogen production plant utilizing nuclear heat has never been built in the world, little attention has been given to the establishment of a safety design for such system including the High Temperature engineering Test Reactor (HTTR). In the present study, requirements in order to design, construct and operate hydrogen production plants under conventional chemical plant standards are identified. In addition, design considerations for safety design of nuclear facility are suggested. Furthermore, feasibility of proposed safety design and design considerations are evaluated. (author)

  11. Design of Vertical Wall Caisson Breakwaters using Partial Safety Factors

    DEFF Research Database (Denmark)

    Burcharth, H. F.; Sørensen, John Dalsgaard

    1999-01-01

    The paper presents a new system for implementation of target reliability in caisson breakwater designs by means of partial safety factors. The development of the system is explained, and tables of partial safety factors are presented for important overall stability failure modes related to caisson...

  12. A German perspective on advances in safety standards and regulations

    International Nuclear Information System (INIS)

    Berg, H.P.; Herttrich, P.M.

    1993-01-01

    At present, different proposals for evolutionary or innovative reactors are under consideration. Therefore, it is necessary that the regulators give guidance on the required safety characteristics of future designs of nuclear power plants. On the one hand, existing regulations have to be updated according to the current state of science and technology. Best available and adequately approved technology has to be used as a yardstick for the acceptability of future basic design features. On the other hand, potential safety features of innovative or revolutionary designs must be considered as serious competitors and potential technical solutions taking the state of maturity of the concepts, the extent of practical experience and the level of effort needed for realization into due account. On this background, recent developments of the Atomic Energy Act, of safety regulations and investigations of requirements for future designs in the Federal Republic of Germany and current projects of international cooperation are presented. (author)

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

    International Nuclear Information System (INIS)

    Duckwitz, Noel

    2011-01-01

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

  14. Development of safety analysis technology for LMR

    International Nuclear Information System (INIS)

    Hahn, Do Hee; Kwon, Y. M.; Kim, K. D.

    2000-05-01

    The analysis methodologies as well as the analysis computer code system for the transient, HCDA, and containment performance analyses, which are required for KALIMER safety analyses, have been developed. The SSC-K code has been developed based on SSC-L which is an analysis code for loop type LMR, by improving models necessary for the KALIMER system analysis, and additional models have been added to the code. In addition, HCDA analysis model has been developed and the containment performance analysis code has been also improved. The preliminary basis for the safety analysis has been established, and the preliminary safety analyses for the key design features have been performed. In addition, a state-of-art analysis for LMR PSA and overseas safety and licensing requirements have been reviewed. The design database for the systematic management of the design documents as well as design processes has been established as well

  15. Development of safety analysis technology for LMR

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, Do Hee; Kwon, Y. M.; Kim, K. D. [and others

    2000-05-01

    The analysis methodologies as well as the analysis computer code system for the transient, HCDA, and containment performance analyses, which are required for KALIMER safety analyses, have been developed. The SSC-K code has been developed based on SSC-L which is an analysis code for loop type LMR, by improving models necessary for the KALIMER system analysis, and additional models have been added to the code. In addition, HCDA analysis model has been developed and the containment performance analysis code has been also improved. The preliminary basis for the safety analysis has been established, and the preliminary safety analyses for the key design features have been performed. In addition, a state-of-art analysis for LMR PSA and overseas safety and licensing requirements have been reviewed. The design database for the systematic management of the design documents as well as design processes has been established as well.

  16. Analysis of effect of safety classification on DCS design in nuclear power plants

    International Nuclear Information System (INIS)

    Gou Guokai; Li Guomin; Wang Qunfeng

    2011-01-01

    By analyzing the safety classification for the systems and functions of nuclear power plants based on the general design requirements for nuclear power plants, especially the requirement of availability and reliability of I and C systems, the characteristics of modem DCS technology and I and C products currently applied in nuclear power field are interpreted. According to the requirements on the safety operation of nuclear power plants and the regulations for safety audit, the effect of different safety classifications on DCS design in nuclear power plants is analyzed, by considering the actual design process of different DCS solutions in the nuclear power plants under construction. (authors)

  17. IEEE standard for design qualification of safety systems equipment used in nuclear power generating stations

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    This standard is written to serve as a general standard for qualification of all types of safety systems equipment, mechanical and instrumentation as well as electrical. It also establishes principles and procedures to be followed in preparing specific safety systems equipment standards. Guidance for qualifying specific safety systems equipment may be found in various specific equipment qualification standards that are now available or are being prepared. It is required that safety systems equipment in nuclear power generating stations meet or exceed its performance requirements throughout its installed life. This is accomplished by a disciplined program of design qualification and quality assurance of design, production, installation, maintenance and surveillance. This standard is for the design qualification section of the program only. Design qualification is intended to demonstrate the capability of the equipment design to perform its safety function(s) over the expected range of normal, abnormal, design basis event, post design basis event, and in-service test conditions. Inherent to design qualification is the requirement for demonstration, within limitations afforded by established technical state-of-the-art, that in-service aging throughout the qualified life established for the equipment will not degrade safety systems equipment from its original design condition to the point where it cannot perform its required safety function(s), upon demand. The above requirement reflects the primary role of design qualification to provide reasonable assurance that design- and age-related common failure modes will not occur during performance of safety function(s) under postulated service conditions

  18. License Application Design Selection Feature Report: Aging and Blending

    International Nuclear Information System (INIS)

    Coltoni, B.; Anderson, M.J.

    1999-01-01

    The purpose of this document is to evaluate the concepts of Aging and Blending for waste sent to the Monitored Geologic Repository (MGR). These design features are based on pre-emplacement treatment of the waste stream. The envelope of the analysis has been performed under the direction of the License Application Design Selection Team (LADST), which advocated utilizing the Viability Assessment (VA) repository design (DOE 1998c) as the basis. Therefore, this evaluation attempts to modify the VA design only to the extent that Aging and Blending can be accomplished. This modified VA design will be contrasted to the VA Design and the difference in design, costs, and performance will be presented

  19. Basic concept of fuel safety design and assessment for sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Nakae, Nobuo; Baba, Toshikazu; Kamimura, Katsuichiro

    2013-03-01

    'Philosophy in Safety Evaluation of Fast Breeder Reactors' was published as a guideline for safety design and safety evaluation of Sodium-Cooled Fast Reactor in Japan. This guideline points out that cladding creep and swelling due to internal pressure should be taken into account since the fuel is used under high temperature and high burnup, and that fuel assembly deformation and the prevention from coolant channel blockage should be taken into account in viewpoints of nuclear and thermal hydraulic design. However, the requirements including their criteria and evaluation items are not described. Two other domestic guidelines related to core design are applied for fuel design of fast reactor, but the description is considered to not be enough to practically use. In addition, technical standard for nuclear fuel used in power reactors is also applied for fuel inspection. Therefore, the technical standard and guideline for fuel design and safety evaluation are considered to be very important issue for nuclear safety regulation. This document has been developed according to the following steps: The guidelines and the technical standards, which are prepared in foreign countries and international organization, were reviewed. The technical background concerning fuel design and safety evaluation for fast reactor was collected and summarized in the world wide scale. The basic concept of fuel safety design and assessment for sodium-cooled fast reactor was developed by considering a wide range of views of the specialists in Japan. In order to discuss the content with foreign specialists IAEA Consultancy Meetings have been held on January, 2011 and January, 2012. The participants of the meeting came from USA, UK, EC, India, China and South Korea. The specialists of IAEA and JNES were also joined. Although this document is prepared for application to 'Monju'(prototype LMFR), it may be applied to experimental, demonstration and commercial types of LMFR after revising it by taking

  20. 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