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Sample records for safety basis upgrade

  1. 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

  2. Westinghouse Hanford Company safety analysis reports and technical safety requirements upgrade program

    International Nuclear Information System (INIS)

    Busche, D.M.

    1995-09-01

    During Fiscal Year 1992, the US Department of Energy, Richland Operations Office (RL) separately transmitted the following US Department of Energy (DOE) Orders to Westinghouse Hanford Company (WHC) for compliance: DOE 5480.21, ''Unreviewed Safety Questions,'' DOE 5480.22, ''Technical Safety Requirements,'' and DOE 5480.23, ''Nuclear Safety Analysis Reports.'' WHC has proceeded with its impact assessment and implementation process for the Orders. The Orders are closely-related and contain some requirements that are either identical, similar, or logically-related. Consequently, WHC has developed a strategy calling for an integrated implementation of the three Orders. The strategy is comprised of three primary objectives, namely: Obtain DOE approval of a single list of DOE-owned and WHC-managed Nuclear Facilities, Establish and/or upgrade the ''Safety Basis'' for each Nuclear Facility, and Establish a functional Unreviewed Safety Question (USQ) process to govern the management and preservation of the Safety Basis for each Nuclear Facility. WHC has developed policy-revision and facility-specific implementation plans to accomplish near-term tasks associated with the above strategic objectives. This plan, which as originally submitted in August 1993 and approved, provided an interpretation of the new DOE Nuclear Facility definition and an initial list of WHC-managed Nuclear Facilities. For each current existing Nuclear Facility, existing Safety Basis documents are identified and the plan/status is provided for the ISB. Plans for upgrading SARs and developing TSRs will be provided after issuance of the corresponding Rules

  3. Safety upgrading at PAKS Nuclear Power Plant

    International Nuclear Information System (INIS)

    Bajsz, J.; Elter, J.

    2000-01-01

    The operation of Paks NPP has reached its half time. Until this time the plant fulfilled expectations raised before its construction: the four units have produced safely and reliably more than 200 TWh electricity. The production of the plant has been at the stable level since its construction and has provided 43-38 % of electricity consumed in Hungary. The annual production is around 14 TWh, which means a load factor higher than 85 %. Safety upgrading activities [1] at Paks had started in the late eighties, when the commissioning work of units 3 and 4 were carried out. That time the main emphases were put to lessons learned of the TMI and Chernobyl accidents. The international reviews hosted by our plant widened our review's scope. To systematize our approach a complete safety review, the AGNES (Advanced General Safety and New Evaluation of Safety) project was started in 1991. The goal of the project was to evaluate to what extent Paks NPP satisfied the current international safety expectations and to help in determining the priorities for safety enhancement and upgrading measures. The project completed in 1994 ranked our safety upgrading measures by safety significance, which became a basis for technical design work and financial scheduling. The other important outcome of the AGNES project was the introduction the Periodical Safety Review regime by our nuclear authority. These periodical reviews held after 10 years of operation offer the possibility - and obligation for the licensee - to perform a comprehensive assessment of the safety of the plant, to evaluate the integral effects of changes of circumstances happened during the review period. The goal of these reviews is to deal with cumulative effects of NPP ageing, modifications, operating experience and technical developments aimed at ensuring a high level of safety throughout plant service life. The execution of our safety-upgrading program is well advancing. For the whole program from 1996 to 2002 250

  4. VVER 1000-NPP Temelin safety upgrading

    International Nuclear Information System (INIS)

    Fleischhans, J.; Ubra, O.

    1995-01-01

    A modernisation program upgrading Temelin plant to meet internationally adopted standard has been implemented during plant design and construction phases. The initial Czech-Russian design (primary system was of Russian design, secondary system was of Czech design) has been extensively modified and adapted to present western safety criteria and operational requirements. The goals are to achieve a high level of safety, reliability, availability and load-following ability. The load-following ability and response to grid frequency changes are very important for the Czech Republic, since the nuclear capacity represents a high proportion of the overall electrical system there. On the basis of IAEA OSART missions and Halliburton NUS audit results and in compliance with recommendations of The State Office for Nuclear Safety, Czech Power Company and Czech scientists and researchers a modernisation program project for Temelin has been carried out. It includes three main groups of VVER1000 MW unit innovations: - Modernization and upgrading of the safety and control systems. - Fuel replacement and modification of the reactor core. - Innovation of some components of the primary and secondary systems. The tenders for instrumentation and control system, nuclear fuel, diagnostic system and radiation monitoring system were issued to the world-well known suppliers. The US company Westinghouse Electric >Corporation (WEC) was selected to submit contract for the delivery of instrumentation and control system primary side diagnostic system and for the delivery of nuclear fuel. The contract was signed in 1993

  5. The concept of underground nuclear heat and power plants (UNHPP) of upgraded safety, developed on the basis of ship-building technologies

    International Nuclear Information System (INIS)

    Pashin, V.M.; Petrov, Eh.L.; Shalik, G.P.; Khazov, B.S.; Malyshev, S.P.

    1996-01-01

    A concept of underground nuclear heat and power plants (UNHPP) of upgraded safety on the basis of ship-building technologies is considered, in which the priority is set to population security and environmental protection. Ways of realization of ziro radiation risk for the population residing in a close vicinity of UNHPP are substantiated. basic principles of the concept are formulated which envisage the use of ship propulsion reactor facilities that have been multiply tested in operation. The sources of economic competitiveness of UNHPPs, as compared with the existing NPPs, are shown

  6. Shortened outage duration and increased safety with head assembly upgrade packages

    International Nuclear Information System (INIS)

    Leanne, M.; Lisien, P.E.; Plute, K.; Duran, J.

    2007-01-01

    To significantly reduce outage critical path duration and personnel radiation exposure, and to increase personnel safety, Westinghouse Electric Co., LLC has designed and installed upgrades to the existing head assemblies of operating pressurized water reactors. These upgrades are known as Head Assembly Upgrade Packages (HAUPs) or Simplified Head Assemblies (SHAs). Custom configurations are created from a set of standard elements to optimize the design for each unique containment, head assembly configuration, and licensing basis. Two primary options are available for implementation: a full HAUP or targeted component and system upgrades. Plants may achieve much of the outage savings, dose reduction, and safety improvements even with a more limited hardware scope. A range of improvements can be offered from integral missile shields, to redesigned duct work, radiation shields, and cable layout and connection optimization. The hardware changes are customized to target the scope that adds the most value for a given plant. While combining upgrades with a reactor vessel head (RVH) replacement adds some flexibility, it is not necessary. Some plants have chosen to implement targeted upgrades prior to a replacement RVH outage and then complete the remainder of the full HAUP during the replacement RVH outage. Three-dimensional computer aided design tools are used in the conceptual and detailed design phases to identify and avoid interferences between existing and replacement plant components. State-of-the-art computational fluid dynamics models for control drive mechanism (CDM) cooling systems are used to demonstrate the ability to maintain or improve the original design performance while greatly simplifying the disassembly/re-assembly activities. Likewise, state-of-the-art finite element analysis methods allow optimization of structural components while meeting code limits for design basis accident conditions. (authors)

  7. 340 Waste Handling Facility interim safety basis

    International Nuclear Information System (INIS)

    Bendixsen, R.B.

    1995-01-01

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

  8. Safety upgrades to the NRU research reactor

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  9. Safety upgrades of NPPs in Slovakia

    International Nuclear Information System (INIS)

    Ziakova, M.

    2013-01-01

    In this presentation Mrs. Marta Ziakova (Chairperson od Nuclear Regulatory Authority of the Slovak Republic) deals with the safety upgrades of nuclear power plants and radioactive wastes technologies in the Slovak Republic. Safety improvement activities are reviewed.

  10. Safety system upgrades to a research reactor: A regulatory perspective

    International Nuclear Information System (INIS)

    Lamarre, G.B.; Martin, W.G.

    2003-01-01

    The NRU (National Research Universal) reactor, located at the Chalk River Laboratories of Atomic Energy of Canada Limited (AECL), first achieved criticality November 3, 1957. AECL continues to operate NRU for research to support safety and reliability studies for CANDU reactors and as a major supplier of medical radioisotopes. Following a detailed systematic review and assessment of NRU's design and the condition of its primary systems, AECL formally notified the Canadian Nuclear Safety Commission's (CNSC) predecessor - the Atomic Energy Control Board - in 1992 of its intention to upgrade NRU's safety systems. AECL proposed seven major upgrades to provide improvements in shutdown capability, heat removal, confinement, and reactor monitoring, particularly during and after a seismic event. From a CNSC perspective, these upgrades were necessary to meet modern safety standards. From the start of the upgrades project, the CNSC provided regulatory oversight aimed at ensuring that AECL maintained a structured approach to the upgrades. The elements of the approach include, but are not limited to, the determination of project milestones and target dates; the formalization of the design process and project quality assurance requirements; the requirements for updated documentation, including safety reports, safety notes and commissioning reports; and the approval and authorization process. This paper details, from a regulatory perspective, the structured approach used in approving the design, construction, commissioning and subsequent operation of safety system upgrades for an existing and operating research reactor, including the many challenges faced when attempting to balance the requirements of the upgrades project with AECL's need to keep NRU operating to meet its important research and production objectives. (author)

  11. Safety analysis report upgrade program at the Plutonium Facility, Los Alamos National Laboratory

    International Nuclear Information System (INIS)

    Pan, P.Y.

    1993-01-01

    Plutonium research and development activities have resided at the Los Alamos National Laboratory (LANL) since 1943. The function of the Plutonium Facility (PF-4) has been to perform basic special nuclear materials research and development and to support national defense and energy programs. The original Final Safety Analysis Report (FSAR) for PF-4 was approved by DOE in 1978. This FSAR analyzed design-basis and bounding accidents. In 1986, DOE/AL published DOE/AL Order 5481.1B, ''Safety Analysis and Review System'', as a requirement for preparation and review of safety analyses. To meet the new DOE requirements, the Facilities Management Group of the Nuclear Material Technology Division submitted a draft FSAR to DOE for approval in April 1991. This draft FSAR analyzed the new configurations and used a limited-scope probabilistic risk analysis for accident analysis. During the DOE review of the draft FSAR, DOE Order 5480.23 ''Nuclear Safety Analysis Reports'', was promulgated and was later officially released in April 1992. The new order significantly expands the scope, preparation, and maintenance efforts beyond those required in DOE/AL Order 5481.1B by requiring: description of institutional and human-factor safety programs; clear definitions of all facility-specific safety commitments; more comprehensive and detailed hazard assessment; use of new safety analysis methods; and annual updates of FSARs. This paper describes the safety analysis report (SAR) upgrade program at the Plutonium Facility in LANL. The SAR upgrade program is established to meet the requirements in DOE Order 5480.23. Described in this paper are the SAR background, authorization basis for operations, hazard classification, and technical program elements

  12. Nitric Acid Revamp and Upgrading of the Alarm & Protection Safety System at Petrokemija, Croatia

    Directory of Open Access Journals (Sweden)

    Hoško, I.

    2012-04-01

    Full Text Available Every industrial production, particularly chemical processing, demands special attention in conducting the technological process with regard to the security requirements. For this reason, production processes should be continuously monitored by means of control and alarm safety instrumented systems. In the production of nitric acid at Petrokemija d. d., the original alarm safety system was designed as a combination of an electrical relay safety system and transistorized alarm module system. In order to increase safety requirements and modernize the technological process of nitric acid production, revamping and upgrading of the existing alarm safety system was initiated with a new microprocessor system. The newly derived alarm safety system, Simatic PCS 7, links the function of "classically" distributed control (DCS and logical systems in a common hardware and software platform with integrated engineering tools and operator interface to meet the minimum safety standards with safety integrity level 2 (SIL2 up to level 3 (SIL3, according to IEC 61508 and IEC 61511. This professional paper demonstrates the methodology of upgrading the logic of the alarm safety system in the production of nitric acid in the form of a logical diagram, which was the basis for a further step in its design and construction. Based on the mentioned logical diagram and defined security requirements, the project was implemented in three phases: analysis and testing, installation of the safety equipment and system, and commissioning. Developed also was a verification system of all safety conditions, which could be applied to other facilities for production of nitric acid. With the revamped and upgraded interlock alarm safety system, a new and improved safety boundary in the production of nitric acid was set, which created the foundation for further improvement of the production process in terms of improved analysis.

  13. Bohunice Nuclear Power Plant Safety Upgrading Program

    International Nuclear Information System (INIS)

    Toth, A.; Fagula, L.

    1996-01-01

    Bohunice nuclear Power Plant generation represents almost 50% of the Slovak republic electric power production. Due to such high level of commitment to nuclear power in the power generation system, a special attention is given to safe and reliable operation of NPPs. Safety upgrading and operational reliability improvement of Bohunice V-1 NPP was carried out by the Bohunice staff continuously since the plant commissioning. In the 1990 - 1993 period extensive projects were realised. As a result of 'Small Reconstruction of the Bohunice V-1 NPP', the standards of both the nuclear safety and operational reliability have been significantly improved. The implementation of another modifications that will take place gradually during extended refuelling outages and overhauls in the course of 1996 through 1999, is referred to as the Gradual Reconstruction of the Bohunice V-1 Plant. The general goal of the V-1 NPP safety upgrading is the achievement of internationally acceptable level of nuclear safety. Extensive and financially demanding modification process of Bohunice V-2 NPP is likely to be implemented after a completion of the Gradual Reconstruction of the Bohunice V-1 NPP, since the year 1999. With this in mind, a first draft of the strategy of the Bohunice V-2 NPP upgrading program based on Probabilistic Safety assessment consideration was developed. A number of actions with a general effect on Bohunice site safety is evident. All these activities are aimed at reaching the essential objective of Bohunice NPP Management - to ensure a safe, reliable and effective electric energy and heat generation at the Bohunice site. (author)

  14. Safety upgrade at the Leningrad NPP

    International Nuclear Information System (INIS)

    Eperin, A.P.

    1996-01-01

    The LNPP was developed according to the standards of early 70's but, at the same time, during the whole period of operation, the Plant equipment, technological, automatic and control and protection systems were upgraded with regard to changing safety and reliability requirements. Main steps taken during the backfitting stage to improve the reliability and safety of LNPP equipment and systems are discussed

  15. Safety upgrading program in NPP Mochovce

    International Nuclear Information System (INIS)

    Baumeister, P.

    1999-01-01

    EMO interest is to operate only nuclear power plants with high standards of nuclear safety. This aim EMO declare on preparation completion and commissioning of Mochovce Nuclear Power Plant. Wide co-operation of our company with International Atomic Energy Agency and west European Inst.ions and companies has been started with aim to fulfil the nuclear safety requirements for Mochovce NPP. Set of 87 safety measures was implemented at Mochovce Unit 1 and is under construction at Unit 2. Mochovce NPP approach to safety upgrading implementation is showed on chosen measures. This presentation is focused on the issues category III.(author)

  16. Safety upgrading of Bohunice V1 NPP

    International Nuclear Information System (INIS)

    2001-01-01

    This CD is multimedia presentation of programme safety upgrading of Bohunice V1 NPP. It consist of next chapters: (1) Introductory speeches; (2) Nuclear power plant WWER 440; (3) Safety improvement; (4) Bohunice Nuclear power plants subsidiary; (5) Siemens; (6) REKON; (7) VUJE Trnava, Inc. - Engineering, Design and Research Organisation; (8) Album

  17. INPP safety upgrade programme. Accomplishments and progress

    International Nuclear Information System (INIS)

    Vaisnys, P.

    1996-01-01

    The safety upgrade programme consists of the following parts: Post Chernobyl immediate modifications undertaken to remove obvious deficiencies discovered in the course of analysis of the main causes of Chernobyl accident; Modifications to remove gaps in safety provision discovered as a result of safety assessment investigations; Modifications to remove evident discrepancies in respect to internationally accepted standard. As it follows from above the deep safety investigations were undertaken to put their findings into concrete improvement programme

  18. Safety Basis Report

    International Nuclear Information System (INIS)

    R.J. Garrett

    2002-01-01

    As part of the internal Integrated Safety Management Assessment verification process, it was determined that there was a lack of documentation that summarizes the safety basis of the current Yucca Mountain Project (YMP) site characterization activities. It was noted that a safety basis would make it possible to establish a technically justifiable graded approach to the implementation of the requirements identified in the Standards/Requirements Identification Document. The Standards/Requirements Identification Documents commit a facility to compliance with specific requirements and, together with the hazard baseline documentation, provide a technical basis for ensuring that the public and workers are protected. This Safety Basis Report has been developed to establish and document the safety basis of the current site characterization activities, establish and document the hazard baseline, and provide the technical basis for identifying structures, systems, and components (SSCs) that perform functions necessary to protect the public, the worker, and the environment from hazards unique to the YMP site characterization activities. This technical basis for identifying SSCs serves as a grading process for the implementation of programs such as Conduct of Operations (DOE Order 5480.19) and the Suspect/Counterfeit Items Program. In addition, this report provides a consolidated summary of the hazards analyses processes developed to support the design, construction, and operation of the YMP site characterization facilities and, therefore, provides a tool for evaluating the safety impacts of changes to the design and operation of the YMP site characterization activities

  19. Safety Basis Report

    Energy Technology Data Exchange (ETDEWEB)

    R.J. Garrett

    2002-01-14

    As part of the internal Integrated Safety Management Assessment verification process, it was determined that there was a lack of documentation that summarizes the safety basis of the current Yucca Mountain Project (YMP) site characterization activities. It was noted that a safety basis would make it possible to establish a technically justifiable graded approach to the implementation of the requirements identified in the Standards/Requirements Identification Document. The Standards/Requirements Identification Documents commit a facility to compliance with specific requirements and, together with the hazard baseline documentation, provide a technical basis for ensuring that the public and workers are protected. This Safety Basis Report has been developed to establish and document the safety basis of the current site characterization activities, establish and document the hazard baseline, and provide the technical basis for identifying structures, systems, and components (SSCs) that perform functions necessary to protect the public, the worker, and the environment from hazards unique to the YMP site characterization activities. This technical basis for identifying SSCs serves as a grading process for the implementation of programs such as Conduct of Operations (DOE Order 5480.19) and the Suspect/Counterfeit Items Program. In addition, this report provides a consolidated summary of the hazards analyses processes developed to support the design, construction, and operation of the YMP site characterization facilities and, therefore, provides a tool for evaluating the safety impacts of changes to the design and operation of the YMP site characterization activities.

  20. 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

  1. The Bohunice NPP V-1 units nuclear safety upgrading

    International Nuclear Information System (INIS)

    Mlcuch, M.

    2000-01-01

    Safety upgrading and operational reliability improvement was carried out by the Bohunice NPP V-1 staff continuously since the plant commissioning. By now, more than 1200 minor or major modifications have been implemented, either by the NPP maintenance staff or by the contractors. Based on findings of safety assessment missions invited by Bohunice NPP in 1990 - 1991, the Czecho-slovak Nuclear Regulatory Authority (CSKAE) issued the decision No. 5/91 of 81 safety upgrading measures to be taken in different areas. These improvements are referred to as the 'Small Reconstruction of the Bohunice V-1 NPP'. Realization of measures during Small reconstruction of the Bohunice NPP V-1 became a power plant, which further operation is acceptable from safety point of view, but it is also necessary further safety improvement. During the period of the Small Reconstruction the development of a Safety Report for the Gradual reconstruction has been completed. Based on this report the SR Nuclear Regulatory Authority issued the Decision No. 1/94, in which requires 59 upgrading measures in different areas to be addressed. The development of Basic Engineering of the Gradual Reconstruction has been contracted to the Siemens AG. Implementation of safety measures are provided through contract with the consortium REKON (which consists of Siemens AG company and Nuclear Power Plants Research Institute Trnava) and other Czech, Russian and Slovak companies. The Gradual Reconstruction of Bohunice NPP V-1 will be finished in 2000. By implementation of the measures carried out during Gradual Reconstruction achievement of an internationally acceptable nuclear safety level will be reached. (author)

  2. Bohunice V-1. Review of safety upgrading and operating experiences

    International Nuclear Information System (INIS)

    Korec, J.; Kuschel, D.

    2000-01-01

    The Bohunice site in the Slovak Republic has two Russian-designed twin-unit nuclear power plants, one equipped with reactors of the WWER 440/230 type, the other with type WWER 440/213 reactors. Two older units (V-1) started commercial operation in late 1978 and 1980 respectively and have been supplying electricity to the national grid since that time without any events that could have degraded plant safety level. In the period prior to 1990 the utility Slovenske Elektrarne (S.E.) performed extensive modifications and upgrades to the original design of the two older units V-1 NPP. Furthermore, significant steps in safety improvement for Bohunice NPP V-1 have been made since 1990. Following the political restructuring of the former Czechoslovakia and the country's new open-door policy towards western organizations, several international expert missions were focused on evaluation of Bohunice NPP safety status level and operational reliability, particularly targeting the two older units. Based on recommendations of individual expert missions and complementary deterministic and probabilistic safety analyses performed by S.E., the Czechoslovak Nuclear Regulatory Authority issued the Resolution No. 5/91 defining 81 measures concerning further safety and reliability improvement of Bohunice V1 .A range of short-term and long-term upgrades was prioritised in terms of importance to plant safety and work to implement these measures commenced in the early nineties. During the 'Small Reconstruction' from 1991 to 1993 some of the short term upgrading measures were realized to eliminate the most serious safety deficits, thus to achieve a significant reduction in core damage frequency and a major improvement in confinement integrity. In this paper and presentation the goals of the gradual reconstruction project, basic engineering, detailed engineering and realization, last major stage of Unit 2 upgrade, as well as final stage of Unit 1 upgrade in early 2000 are presented

  3. Upgrading safety systems of industrial irradiation facilities

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  4. Upgrading safety systems of industrial irradiation facilities

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

  5. Armenian nuclear power plant: US NRC assistance programme for seismic upgrade and safety analysis

    International Nuclear Information System (INIS)

    Simos, N.; Perkins, K.; Jo, J.; Carew, J.; Ramsey, J.

    2003-01-01

    This paper summarizes the U.S. Nuclear Regulatory Commission's (US NRC) technical support program activities associated with the Armenian Nuclear Power Plant (ANPP) safety upgrade. The US NRC program, integrated within the overall IAEA-led initiative for safety re-evaluation of the WWER plants, has as its main thrust the technical support to the Armenian Nuclear Regulatory Authority (ANRA) through close collaboration with the scientific staff at Brookhaven National Laboratory (BNL). Several major technical areas of support to ANRA form the basis of the NRC program. These include the seismic re-evaluation and upgrade of the ANPP, safety evaluation of critical systems, and the generation of the Safety Analysis Report (SAR). Specifically, the seismic re-evaluation of the ANPP is part of a broader activity that involves the re-assessment of the seismic hazard at the site, the identification of the Safe Shutdown Equipment at the plant and the evaluation of their seismic capacity, the detailed modeling and analysis of the critical facilities at ANPP, and the generation of the Floor Response Spectra (FRS). Based on the new spectra that incorporate all new findings (hazard, site soil, structure, etc.), the overall capacity of the main structures and the seismic capacity of the critical systems are being re-evaluated. In addition, analyses of critical safe shutdown systems and safe shutdown processes are being performed to ensure both the capabilities of the operating systems and the enhancement of safety due to system upgrades. At present, one of the principal goals of the US NRC's regulatory assistance activities with ANRA is enhancing ANRA's regulatory oversight of high-priority safety issues (both generic and plant-specific) associated with operation of the ANPP. As such, assisting ANRA in understanding and assessing plant-specific seismic and other safety issues associated with the ANPP is a high priority given the ANPP's being located in a seismically active area

  6. South Ukraine NPP: Safety improvements through Plant Computer upgrade

    International Nuclear Information System (INIS)

    Brenman, O.; Chernyshov, M. A.; Denning, R. S.; Kolesov, S. A.; Balakan, H. H.; Bilyk, B. I.; Kuznetsov, V. I.; Trosman, G.

    2006-01-01

    This paper summarizes some results of the Plant Computer upgrade at the Units 2 and 3 of South Ukraine Nuclear Power Plant (NPP). A Plant Computer, which is also called the Computer Information System (CIS), is one of the key safety-related systems at VVER-1000 nuclear plants. The main function of the CIS is information support for the plant operators during normal and emergency operational modes. Before this upgrade, South Ukraine NPP operated out-of-date and obsolete systems. This upgrade project wax founded by the U.S. DOE in the framework of the International Nuclear Safety Program (INSP). The most efficient way to improve the quality and reliability of information provided to the plant operator is to upgrade the Human-System Interface (HSI), which is the Upper Level (UL) CIS. The upgrade of the CIS data-acquisition system (DAS), which is the Lower Level (LL) CIS, would have less effect on the unit safety. Generally speaking, the lifetime of the LL CIS is much higher than one of the UL CIS. Unlike Plant Computers at the Western-designed plants, the functionality of the WER-1000 CISs includes a control function (Centralized Protection Testing) and a number of the plant equipment monitoring functions, for example, Protection and Interlock Monitoring and Turbo-Generator Temperature Monitoring. The new system is consistent with a historical migration of the format by which information is presented to the operator away from the traditional graphic displays, for example, Piping and Instrument Diagrams (P and ID's), toward Integral Data displays. The cognitive approach to information presentation is currently limited by some licensing issues, but is adapted to a greater degree with each new system. The paper provides some lessons learned on the management of the international team. (authors)

  7. Just in Time DSA-The Hanford Nuclear Safety Basis Strategy

    Energy Technology Data Exchange (ETDEWEB)

    Olinger, S. J.; Buhl, A. R.

    2002-02-26

    The U.S. Department of Energy, Richland Operations Office (RL) is responsible for 30 hazard category 2 and 3 nuclear facilities that are operated by its prime contractors, Fluor Hanford Incorporated (FHI), Bechtel Hanford, Incorporated (BHI) and Pacific Northwest National Laboratory (PNNL). The publication of Title 10, Code of Federal Regulations, Part 830, Subpart B, Safety Basis Requirements (the Rule) in January 2001 imposed the requirement that the Documented Safety Analyses (DSA) for these facilities be reviewed against the requirements of the Rule. Those DSA that do not meet the requirements must either be upgraded to satisfy the Rule, or an exemption must be obtained. RL and its prime contractors have developed a Nuclear Safety Strategy that provides a comprehensive approach for supporting RL's efforts to meet its long term objectives for hazard category 2 and 3 facilities while also meeting the requirements of the Rule. This approach will result in a reduction of the total number of safety basis documents that must be developed and maintained to support the remaining mission and closure of the Hanford Site and ensure that the documentation that must be developed will support: compliance with the Rule; a ''Just-In-Time'' approach to development of Rule-compliant safety bases supported by temporary exemptions; and consolidation of safety basis documents that support multiple facilities with a common mission (e.g. decontamination, decommissioning and demolition [DD&D], waste management, surveillance and maintenance). This strategy provides a clear path to transition the safety bases for the various Hanford facilities from support of operation and stabilization missions through DD&D to accelerate closure. This ''Just-In-Time'' Strategy can also be tailored for other DOE Sites, creating the potential for large cost savings and schedule reductions throughout the DOE complex.

  8. Just in Time DSA-The Hanford Nuclear Safety Basis Strategy

    International Nuclear Information System (INIS)

    Olinger, S. J.; Buhl, A. R.

    2002-01-01

    The U.S. Department of Energy, Richland Operations Office (RL) is responsible for 30 hazard category 2 and 3 nuclear facilities that are operated by its prime contractors, Fluor Hanford Incorporated (FHI), Bechtel Hanford, Incorporated (BHI) and Pacific Northwest National Laboratory (PNNL). The publication of Title 10, Code of Federal Regulations, Part 830, Subpart B, Safety Basis Requirements (the Rule) in January 2001 imposed the requirement that the Documented Safety Analyses (DSA) for these facilities be reviewed against the requirements of the Rule. Those DSA that do not meet the requirements must either be upgraded to satisfy the Rule, or an exemption must be obtained. RL and its prime contractors have developed a Nuclear Safety Strategy that provides a comprehensive approach for supporting RL's efforts to meet its long term objectives for hazard category 2 and 3 facilities while also meeting the requirements of the Rule. This approach will result in a reduction of the total number of safety basis documents that must be developed and maintained to support the remaining mission and closure of the Hanford Site and ensure that the documentation that must be developed will support: compliance with the Rule; a ''Just-In-Time'' approach to development of Rule-compliant safety bases supported by temporary exemptions; and consolidation of safety basis documents that support multiple facilities with a common mission (e.g. decontamination, decommissioning and demolition [DD and D], waste management, surveillance and maintenance). This strategy provides a clear path to transition the safety bases for the various Hanford facilities from support of operation and stabilization missions through DD and D to accelerate closure. This ''Just-In-Time'' Strategy can also be tailored for other DOE Sites, creating the potential for large cost savings and schedule reductions throughout the DOE complex

  9. Safety basis for selected activities in single-shell tanks with flammable gas concerns. Revision 1

    International Nuclear Information System (INIS)

    Schlosser, R.L.

    1996-01-01

    This is full revision to Revision 0 of this report. The purpose of this report is to provide a summary of analyses done to support activities performed for single-shell tanks. These activities are encompassed by the flammable gas Unreviewed Safety Question (USQ). The basic controls required to perform these activities involve the identification, elimination and/or control of ignition sources and monitoring for flammable gases. Controls are implemented through the Interim Safety Basis (ISB), IOSRs, and OSDs. Since this report only provides a historical compendium of issues and activities, it is not to be used as a basis to perform USQ screenings and evaluations. Furthermore, these analyses and others in process will be used as the basis for developing the Flammable Gas Topical Report for the ISB Upgrade

  10. 10 CFR 830.202 - Safety basis.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 4 2010-01-01 2010-01-01 false Safety basis. 830.202 Section 830.202 Energy DEPARTMENT OF ENERGY NUCLEAR SAFETY MANAGEMENT Safety Basis Requirements § 830.202 Safety basis. (a) The contractor responsible for a hazard category 1, 2, or 3 DOE nuclear facility must establish and maintain the safety basis...

  11. Upgrading of fire safety in nuclear power plants. Proceedings of an International Symposium

    International Nuclear Information System (INIS)

    1998-04-01

    The document includes 40 papers presented at the International Symposium on Upgrading of Fire Safety in Nuclear Power Plants held in Vienna between 18-21 November 1997. The symposium presentations were grouped in 6 sessions: Fire safety reviews (5 papers), Fire safety analysis - Methodology (6 papers), Fire safety analysis - Applications (3 papers), Panel 1 - Identification of deficiencies in fire safety in nuclear power plants - Operational experience and data (7 papers), Panel 2 - Experience based data in fire safety assessment - Fire safety regulations and licensing (7 papers), Upgrading programmes (10 papers), and a closing session (2 papers). A separate abstract was prepared for each paper

  12. Upgrading of fire safety in nuclear power plants. Proceedings of an International Symposium

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-04-01

    The document includes 40 papers presented at the International Symposium on Upgrading of Fire Safety in Nuclear Power Plants held in Vienna between 18-21 November 1997. The symposium presentations were grouped in 6 sessions: Fire safety reviews (5 papers), Fire safety analysis - Methodology (6 papers), Fire safety analysis - Applications (3 papers), Panel 1 - Identification of deficiencies in fire safety in nuclear power plants - Operational experience and data (7 papers), Panel 2 - Experience based data in fire safety assessment - Fire safety regulations and licensing (7 papers), Upgrading programmes (10 papers), and a closing session (2 papers). A separate abstract was prepared for each paper Refs, figs, tabs

  13. Project W-030 safety class upgrade summary report

    International Nuclear Information System (INIS)

    Kriskovich, J.R.

    1998-01-01

    This document presents a summary of safety class criteria for the 241-AY/AZ Tank Farm primary ventilation system upgrade under Project W-030, and recommends acceptance of the system as constructed, based on a review of supporting documentation

  14. Just in Time DSA the Hanford Nuclear Safety Basis Strategy

    Energy Technology Data Exchange (ETDEWEB)

    JACKSON, M.W.

    2002-06-01

    The U.S. Department of Energy, Richland Operations Office (RL) is responsible for 30 hazard category 2 and 3 nuclear facilities that are operated by its prime contractors, Fluor Hanford, Incorporated (FHI), Bechtel Hanford, Incorporated (BHI) and Pacific Northwest National Laboratory (PNNL). The publication of Title 10, Code of Federal Regulations, Part 830, Subpart B, Safely Basis Requirements (the Rule) in January 2001 requires that the Documented Safety Analyses (DSA) for these facilities be reviewed against the requirements of the Rule. Those DSAs that do not meet the requirements must either be upgraded to satisfy the Rule, or an exemption must be obtained. RL and its prime contractors have developed a Nuclear Safety Strategy that provides a comprehensive approach for supporting RL's efforts to meet its long-term objectives for hazard category 2 and 3 facilities while also meeting the requirements of the Rule. This approach will result in a reduction of the total number of safety basis documents that must be developed and maintained to support the remaining mission and closure of the Hanford Site and ensure that the documentation that must be developed will support: Compliance with the Rule; A ''Just-In-Time'' approach to development of Rule-compliant safety bases supported by temporary exemptions; and Consolidation of safety basis documents that support multiple facilities with a common mission (e.g. decontamination, decommissioning and demolition [DD&D], waste management, surveillance and maintenance). This strategy provides a clear path to transition the safety bases for the various Hanford facilities from support of operation and stabilization missions through DD&D to accelerate closure. This ''Just-In-Time'' Strategy can also be tailored for other DOE Sites, creating the potential for large cost savings and schedule reductions throughout the DOE complex.

  15. Just in Time DSA the Hanford Nuclear Safety Basis Strategy

    International Nuclear Information System (INIS)

    JACKSON, M.W.

    2002-01-01

    The U.S. Department of Energy, Richland Operations Office (RL) is responsible for 30 hazard category 2 and 3 nuclear facilities that are operated by its prime contractors, Fluor Hanford, Incorporated (FHI), Bechtel Hanford, Incorporated (BHI) and Pacific Northwest National Laboratory (PNNL). The publication of Title 10, Code of Federal Regulations, Part 830, Subpart B, Safely Basis Requirements (the Rule) in January 2001 requires that the Documented Safety Analyses (DSA) for these facilities be reviewed against the requirements of the Rule. Those DSAs that do not meet the requirements must either be upgraded to satisfy the Rule, or an exemption must be obtained. RL and its prime contractors have developed a Nuclear Safety Strategy that provides a comprehensive approach for supporting RL's efforts to meet its long-term objectives for hazard category 2 and 3 facilities while also meeting the requirements of the Rule. This approach will result in a reduction of the total number of safety basis documents that must be developed and maintained to support the remaining mission and closure of the Hanford Site and ensure that the documentation that must be developed will support: Compliance with the Rule; A ''Just-In-Time'' approach to development of Rule-compliant safety bases supported by temporary exemptions; and Consolidation of safety basis documents that support multiple facilities with a common mission (e.g. decontamination, decommissioning and demolition [DD and D], waste management, surveillance and maintenance). This strategy provides a clear path to transition the safety bases for the various Hanford facilities from support of operation and stabilization missions through DD and D to accelerate closure. This ''Just-In-Time'' Strategy can also be tailored for other DOE Sites, creating the potential for large cost savings and schedule reductions throughout the DOE complex

  16. Upgrading the safety assessment of exported nuclear power plants

    International Nuclear Information System (INIS)

    Rosen, M.

    1978-01-01

    An examination of the safety aspects of exported nuclear power plants demonstrates that additional and somewhat special considerations exist for these plants, and thus that some new approaches may be required to insure their safety. In view of the generally small regulatory staffs of importing countries, suggestions are given for measures which should be taken by the various organizations involved in the export and import of nuclear power facilities to raise the level of the very essential safety assessment. These include the upgrading of the 'export edition' of the traditionally supplied safety documentation by use of a Supplementary Information Report, written specifically for the needs of a smaller and/or less technically qualified staff, which highlights the differences that exist between the facility to be constructed and the supposedly similar reference plant of the supplier country; by improvement of supporting safety documentation to allow for adequate understanding of significant safety parameters; and by attention to the needs of smaller countries in the critical Operating Regulations (Technical Specifications for Operation). Consideration is also given to upgrading the regulatory effort and to the obligations of principal organizations involved with exported nuclear plants, including national and international, for insuring the importing countries' technical readiness and the adequacy of the regulatory effort. Special attention is directed towards the project contract as a means of implementing programmes to achieve these goals. (author)

  17. Upgrading safety documentation for exported nuclear power plants

    International Nuclear Information System (INIS)

    Rosen, M.

    1978-01-01

    In view of the generally small regulatory staffs of importing countries, suggestions are given for upgrading the ''export edition'' of the traditionally supplied safety documentation by use of a Supplementary Information Report, written specifically for the needs of a smaller and/or less technically qualified staff, which would highlight the differences that exist between the facility to be constructed and the supposedly similar reference plant of the supplier country; by improvement of supporting safety documentation to allow for adequate understanding of significant safety parameters; and by attention to the needs of smaller countries in the critical operating regulations (Technical Specifications for Operation). (author)

  18. Safety philosophy in upgrading the EBR-II plant protection system

    International Nuclear Information System (INIS)

    Sackett, J.I.

    1976-01-01

    The EBR-II plant protection system (PPS) has been substantially modified, upgrading its performance to more fully comply with modern safety philosophy and criteria. The upgrading effort required that the total reactor system be evaluated for possible faults and that a PPS be designed to accommodate them. The result was deletion of a number of existing trip functions and upgrading of others. Particular attention was given to loss of primary pumping power and reactivity insertion events. The design and performance criteria for the PPS has been more firmly established, understanding of the PPS function has been improved and the reactor has been subjected to fewer spurious trips, improving operational reliability

  19. Performing instrumentation and controls upgrades

    International Nuclear Information System (INIS)

    Kessler, F. M.; Connell, T. J.; Ryan, M. P.

    1992-01-01

    I and C upgrades are comprised of a varying range of content, complexity, expansiveness, and criticality. There are common threads in all upgrades which can be simplified by the development of a long term I and C upgrade plan. The development of a such a plan can establish effective ground rules for upgrades, large and small. It can be the basis from which to begin an upgrade evaluation and the standard which is used to compare the degree of compliance of any upgrade regarding the plan or to define the differences from the plan and an individual upgrade. Primary motivation for I and C upgrades are obsolescence and unavailability of spare parts. Numerous other areas of consideration are also involved in an upgrade. Today's technology results in most upgrades largely or totally utilizing digital equipment. The use of digital equipment is fairly new in many I and C applications and requires an elaborate evaluation from functional, qualification, operational, and licensing perspectives as well as others. A well defined upgrade plan developed as a basis for I and C upgrades is a significant start to ensuring an effective upgrade process. Properly developed and implemented, the plan will support I and C upgrade efforts to ensure that the intricacies associated with such tasks eliminate the existing problems which require the upgrade to be performed. The upgrade plan also results in ensuring the maximum benefit from all perspectives of the plant enhancements being carried out and considered for future implementation. Instrumentation and controls aging and replacement are issues of growing importance due to the potential for significant impact on plant operation and efficiency. Obsolescence and unavailability of spare parts are major drivers towards evaluating the cost benefits of upgrading current equipment. In addition to these two primary factors, the advantages of utilizing digital equipment have also become of prime importance when evaluating instrumentation and

  20. Reactivity initiated accident analyses for the safety assessment of upgraded JRR-3

    International Nuclear Information System (INIS)

    Harami, Taikan; Uemura, Mutsumi; Ohnishi, Nobuaki

    1984-08-01

    JRR-3, currently a heavy water moderated and cooled 10 MW reactor, is to be upgraded to a light water moderated and cooled, heavy water reflected 20 MW reactor. This report describes the analytical results of reactivity initiated accidents for the safety assessment of upgraded JRR-3. The following five cases have been selected for the assessment; (1) uncontrolled control rod withdrawal from zero power, (2) uncontrolled control rod withdrawal from full power, (3) removal of irradiation samples, (4) increase of primary coolant flow, (5) failure of heavy water tank. Parameter studies have been made for each of the above cases to cover possible uncertainties. All analyses have been made by a computer code EUREKA-2. The results show that the safety criteria for upgraded JRR-3 are all met and the adequacy of the design is confirmed. (author)

  1. 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

  2. Upgrading nuclear safety and security infrastructure in Yemen

    International Nuclear Information System (INIS)

    Bahran, M.Y.

    2007-01-01

    In 1999 the National Atomic Energy Commission of the Republic of Yemen was established with an emphasis on building Radiation Protection Infrastructure suitable for Yemen and in accordance with international standards. Since then, Yemen arguably has built one of the best such systems in the region if not in the world with respect to the country's needs. This system is going to be upgraded from Radiological Safety and Security system to a Nuclear Safety and Security system. This is to be done as a prerequisite to any further development in the Peaceful Applications of Nuclear Energy, particularly Nuclear Power and Desalination. (author)

  3. Approach of seismic upgrading in Kashiwazaki-Kariwa Nuclear Power Plant

    International Nuclear Information System (INIS)

    Sato, Hitoshi

    2009-01-01

    Because guide for reviewing seismic design of nuclear power reactor facilities was reworked in 2006, we formulated new Design Base Seismic Motion Ss, and we are doing evaluation of seismic safety (back-check). In Japan, depending on aseismatic importance, equipments are classified into S-class, B-class and C-class. For S-class equipments, we evaluate it on the basis of new Ss, and do seismic upgrading. For B-class and C-class equipments, we do seismic upgrading voluntarily on the basis of the experiences of the Niigataken Chuetsu-Oki (NCO) Earthquake. (author)

  4. Safety and upgrading of nuclear power plants with WWER-440/V-230; Voprosy bezopasnosti i modernizatsii AEhS s WWER-440 (V-230)

    Energy Technology Data Exchange (ETDEWEB)

    Melnikov, N [Opytno-Konstruktorskoe Byuro Gidropress, Podol` sk (Russian Federation)

    1996-12-31

    Modernization of WWER-440/V-230 reactors is considered according to the WANO project `Proposal for reconstruction of WWER-440/V-230 reactors` (1990). The main technical measures to be taken relate to: 1. Expanding the considered spectrum of design basis accidents; 2. Provision of first loop integrity; 3. Safety system structure and components reliability enhancement; 4. Reduction o initiating events probability; 5. Enhancement of hermeticity and integrity of the area surrounding the first loop; 6. Fire fighting system upgrade; 7. Radiation control system upgrade; 8. Provision of seismic resistance. Details of the recommended technical steps for some NPPs are presented. 1 tab.

  5. Application of risk assessment in upgrading safety and quality of radiochemical operations

    International Nuclear Information System (INIS)

    Lin, K.H.; Hightower, J.R.; Vaughen, V.C.A.

    1990-01-01

    A Comprehensive Safety Assessment and Upgrade Program (CSAUP) was commenced by the Chemical Technology Division (Chem Tech) at Oak Ridge National Laboratory (ORNL) to achieve excellence in the safety and quality of its operations and condition of its facilities. In the course of conducting CSAUP, a number of issues of concern were identified. The safety risk of these issues has been assessed, and planned actions were prepared for those issues that require corrective actions or improvement/upgrading. The planned actions were evaluated on consideration for the uniqueness of Chem Tech facilities and operations to determine the risks (high, moderate or low) involved by failure to implement the actions. The risk was defined in terms of the frequency and severity of impact. Priority of categorized actions was based on the urgency of the actions. 2 refs., 1 fig., 3 tabs

  6. Upgrading the safety of VVER-440/V-230

    International Nuclear Information System (INIS)

    Kelm, P.; Wenk, W.

    1995-01-01

    Besides measures seeking to restore the status as laid down in the project design, especially backfitting measures must be mentioned which serve to ensure component and pipe integrity. Ensuring component integrity is a problem not only of RPV embrittlement, but also of failure prevention. This aspect was not always taken into account properly. Further activities in the field of component integrity will focus on backing the brittle fracture evaluation of the RPV; qualifying the leak-before-breack criterion for the main pipes and in areas with screwed connections; qualifying the program of in-service inspections. Several operators are currently in the process of drafting backfitting programs. The upgrading measures envisaged must be checked as to their balanced nature. In certain plants, the integrity of the RPV coud turn out to be the weak spot in upgrading measures. As a consquence, concepts seeking to achieve upgrading for long periods of time may differ from one location to the next and even between units. Extensive modifications in systems engineering and building structures generally must be evaluated against the expected improvement in safety of the whole plant. (orig.) [de

  7. Refurbishment and safety upgradation of research reactor Cirus

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    International Nuclear Information System (INIS)

    Brognon, T.

    1993-01-01

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

  9. Upgrading of fire safety in Indian nuclear power plants

    International Nuclear Information System (INIS)

    Agarwal, N.K.

    1998-01-01

    Indian nuclear power programme started with the installation of 2 nos. of Boiling Water Reactor (BWR) at Tarapur (TAPS I and II) of 210 MWe each commissioned in the year 1996. The Pressurized Heavy Water Reactor (PHWR) programme in the country started with the installation of 2x220 MWe stations at Rawatbhatta near Kota (RAPS I and II) in the State of Rajasthan in the sixties. At the present moment, the country has 10 stations in operation. Construction is going on for 4 more units of 220 MWe where as work on two more 500 MWe units is going to start soon. Fire safety systems for the earlier units were engineered as per the state-of-art knowledge available then. The need for review of fire protection systems in the Indian nuclear power plants has also been felt since long almost after Brown's Ferry fire in 1975 itself. Task forces consisting of fire experts, systems design engineers, O and M personnel as well as the Fire Protection engineers at the plant were constituted for each plant to review the existing fire safety provisions in details and highlight the upgradation needed for meeting the latest requirements as per the national as well as international practices. The recommendations made by three such task forces for the three plants are proposed to be reviewed in this paper. The paper also highlights the recommendations to be implemented immediately as well as on long-term basis over a period of time

  10. 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

  11. 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.

  12. 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

  13. Status of High Flux Isotope Reactor (HFIR) post-restart safety analysis and documentation upgrades

    International Nuclear Information System (INIS)

    Cook, D.H.; Radcliff, T.D.; Rothrock, R.B.; Schreiber, R.E.

    1990-01-01

    The High Flux Isotope Reactor (HFIR), an experimental reactor located at the Oak Ridge National Laboratory (ORNL) and operated for the US Department of Energy by Martin Marietta Energy Systems, was shut down in November, 1986 after the discovery of unexpected neutron embrittlement of the reactor vessel. The reactor was restarted in April, 1989, following an extensive review by DOE and ORNL of the HFIR design, safety, operation, maintenance and management, and the implementation of several upgrades to HFIR safety-related hardware, analyses, documents and procedures. This included establishing new operating conditions to provide added margin against pressure vessel failure, as well as the addition, or upgrading, of specific safety-related hardware. This paper summarizes the status of some of the follow-on (post-restart) activities which are currently in progress, and which will result in a comprehensive set of safety analyses and documentation for the HFIR, comparable with current practice in commercial nuclear power plants. 8 refs

  14. Improved reliability, maintainability and safety through elastomer upgrading

    International Nuclear Information System (INIS)

    Wensel, R.; Wittich, K.C.

    1995-01-01

    Equipment in nuclear plants has historically contained whatever elastomer each component supplier traditionally used for corresponding non-nuclear service. The resulting proliferation of elastomer compounds, many of which are far from optimal for the service conditions (e.g., pressure, temperature, radiation, etc.), has multiplied the costs to provide station reliability, maintainability and safety. Cost-effective improvements are being achieved in CANDU plants by upgrading and standardizing on a handful of high performing elastomer compounds. These upgraded materials offer significant gains in service life over the materials they replace (often by factors of 2 or more). This rationalization of elastomer compounds also facilitates the EQ process for safety-related equipment. Detailed test data on aging is currently being generated for these specific elastomers, encompassing the conditions and media (air, water, oil) common in CANDU service. Two key elements characterize this testing. First, each result is specific to the compound used in the test, and second, it is specific to the tested failure mode (e.g., compression set, extrusion, fracture, etc.). Having fewer, but more thoroughly tested compounds, avoids the penalty (associated with poorly characterized materials) of having to replace parts prematurely because of conservatism, while maintaining safe, reliable service. This paper provides an overview of this approach covering: the benefits of compound rationalization; and the how and why of establishing relevant failure criteria; appropriate quality assurance to maintain EQ; procurement, storage and handling guidelines; and monitoring and predicting in-service degradation. (author)

  15. Safety upgrades for NSRRC beamlines in the upcoming top-up operation

    International Nuclear Information System (INIS)

    Liu, Joseph C.; Sheu, R.-J.; Wang, J.-P.; Chen, C.-R.; Chang, F.-D.; Kao, S.-P.

    2006-01-01

    The original beamline shielding of NSRRC was designed for the decay mode operation that safety shutter was closed during injection. The proposed top-up operation that opens safety shutter during top-up injection will introduce additional beam loss scenarios and radiation sources, especially when the injection efficiency needs to be improved. Careful comparison was made to differentiate the radiation doses into beamlines for both operation modes. Detailed evaluation was made to identify the possible inadequacies of the old beamline shielding and safety control procedures. Remedy actions and safety upgrades for each individual beamline were issued to ensure that dose limit of 2 mSv/yr for users can be fulfilled when running top-up operation

  16. Upgrading instrumentation and control systems for plant safety and operation

    International Nuclear Information System (INIS)

    Martin, M.; Prehler, H.J.; Schramm, W.

    1997-01-01

    Upgrading the electrical systems and instrumentation and control systems has become increasingly more important in the past few years for nuclear power plants currently in operation. As the requirements to be met in terms of plant safety and availability have become more stringent in the past few years, Western plants built in the sixties and seventies have been the subject of manifold backfitting and upgrading measures in the past. In the meantime, however, various nuclear power plants are facing much more thorough upgrading phases because of the difficulties in obtaining spare parts for older equipment systems. As digital technology has become widespread in many areas because of its advantages, and as applications are continuously expanding, conventional equipment and systems are losing more and more ground as a consequence of decreasing demand. Merely because of the pronounced decline in demand for conventional electronic components it is possible for equipment manufacturers to guarantee spare parts deliveries for older systems only for specific future periods of time. In addition, one-off manufacture entails high costs in purchases of spare parts. As a consequence of current thinking more and more focusing on availability and economy, upgrading of electrical systems and instrumentation and control systems is becoming a more and more topical question, for older plants even to ensure completion of full service life. (orig.) [de

  17. Krsko NPP Quality Assurance Plan Application to Nuclear Safety Upgrade Projects (PCFV System and PAR System)

    International Nuclear Information System (INIS)

    Biscan, Romeo; Fifnja, Igor

    2014-01-01

    Nuklearna Elektrarna Krsko (NEK) has undertaken Nuclear Safety Upgrade Projects as a safety improvement driven by the lessons learned from the Fukushima-Daiichi Accident. Among other projects, new modification 1008-VA-L Passive Containment Filtered Vent (PCFV) System has been installed which acts as the last barrier minimizing the release of radioactive material into the environment in case of failure of all safety systems, and to insure containment integrity during beyond design basis accidents (BDBA). In addition, modification 1002-GH-L Severe Accident Hydrogen Control System (PAR) has been implemented to prevent and mitigate the consequences of explosive gas generation (hydrogen and carbon monoxide) in case of reactor core melting. To ensure containment integrity for all design basis accidents (DBA) and BDBA conditions, NEK has eliminated existing safety-related electrical recombiners, replaced them with two safety-related passive autocatalytic recombiners (PARs) and added 20 new PARs designed for the BDBA conditions. Krsko NPP Quality Assurance Plan has been applied to Nuclear Safety Upgrade Projects (PCFV System and PAR System) through the following activities: · Internal audit of modification process was performed. · Supplier audits were performed to evaluate QA program efficiency of the main design organization and engineering organizations. · Evaluation and approval of Suppliers were performed. · QA engineer was involved in the review and approval of 1008-VA-L and 1002-GH-L modification documentation (Conceptual Design Package, Design Modification Package, Installation Package, Field Design Change Request, Problem/Deficiency Report, and Final Documentation Package). · Purchasing documentation for modifications 1008-VA-L and 1002-GH-L (technical specifications, purchase orders) has been verified and approved by QA. · QA and QC engineers were involved in oversight of production and testing of the new 1008-VA-L and 1002-GH-L plant components.

  18. The US nuclear safety approach to upgrading the Russian and Ukrainian reactors

    International Nuclear Information System (INIS)

    Baron, S.

    1993-01-01

    Brookhaven National Laboratory reporting to the Department of Energy has the technical and administrative management responsibilities for improving the operational and design safety systems of RBMK and the VVER reactors in Russia and the Ukraine. U.S. experts and industry interact with the Russian/Ukraine designers and operators to jointly develop the detailed requirements for system upgrades. When available, indigenous equipment and materials will be utilized. The construction and installation of most upgrades will be accomplished by Russia and the Ukraine with U.S. support and participation. This will maximize technology transfer, provide funds to U.S. and recipient country industry, and limit the nuclear liability of U.S. industry. (author)

  19. The V-1 NPP and V-2 NPP upgrading

    International Nuclear Information System (INIS)

    1998-01-01

    A brief account of activities in the V-1 NPP and V-2 NPP upgrading as well as maintenance carried out by the Nuclear power plants Jaslovske Bohunice in 1997 is presented. The V-1 NPP applied the so called 'Small Backfitting Programme'covering 81 points of the Czechoslovak Atomic energy Commission Decree No 5/91. Continual upgrading continued after the Backfitting Programme completion with the Safety Report and following Nuclear Regulatory Authority of Slovak Republic (NRA SR) Decrees No 1/94 and 110/94 setting spheres and procedure for adopting and implementation of measures enabling the units to operate further on. Results of expert missions, analyses and assessments of components identified by Basic Engineering became the basis for the development of the Gradual Reconstruction Programme. The Programme outputs underwent economic and probabilistic assessing their contribution to nuclear safety. This process resulted in finalizing the Gradual Reconstruction Programme which started to be implemented in 1996 and will be completed in 1999. It is implemented by the REKON consortium and covers 17 areas including Instrumentation and Control, self-consumption emergency supply, leakage monitoring, emergency core cooling system, seismic reinforcement and radioactivity localisation. Both units will reach internationally acceptable safety standards for the remaining life-time period. The V-2 NPP Upgrading and Safety Enhancement Programme includes results of activities performed in the course of last years to define all important activities leading to enhancement of nuclear safety and performance reliability and effectiveness within the plant life-time period and to establish conditions for extending the life-time of these units for 40 years. The V-2 NPP Upgrading and Safety Enhancement Programme aims to assure safe operation with a probability of the core damages less than 10 -4 /reactor · year

  20. Food Safety and Sanitary Practices of Selected Hotels in Batangas Province, Philippines: Basis of Proposed Enhancement Measures

    Directory of Open Access Journals (Sweden)

    April M. Perez

    2017-02-01

    Full Text Available This study assessed the extent of food safety and sanitary practices of selected hotels in Batangas province as basis of proposed enhancement measures. The study utilized descriptive method to describe food safety and sanitary practices of selected hotels in Batangas province with a total of 8 hotels (256 respondents. Purposive sampling was used in the study. The questionnaires were designed using the provision of the Sanitation Code of the Philippines, validated and finalized to come up with legitimate results. The study showed that there were eight (8 hotel respondents classified as two, three, four star with considerable years of experience and adequate number of employees. The hotels demonstrated the food safety and sanitary practices always in the areas of restaurant, bar service, catering and banquet and room service. The significant pair-wise comparison for restaurant, bar service, catering and banquet and room service shows that 2 star hotels greatly differs. The researcher recommends that the management should maintain high standard of food safety and sanitary practices among its staff, upgrade the food safety and sanitary practices for food safety accreditation, continuous training of the hotel managers/employees on food safety and sanitary practices.

  1. Experience with safety assessment of digital upgrading of IandC in VVER type reactors

    International Nuclear Information System (INIS)

    Wach, D.; Mulka, B.; Schnuerer, G.

    1997-01-01

    The digital upgrading of IandC systems important to safety in WWER type reactors requires a broad expertise in various knowledge fields. The approach of the Institute for safety Technology to the qualification and categorization of safety-critical software systems is highlighted. The role of the Institute in the qualification of the Teleperm XS and the type testing of its components is described. The aspects of the safety assessment of digital IandC systems in WWER type reactors is discussed in some detail. (A.K.)

  2. UPGRADES

    CERN Multimedia

    J. Spalding and D. Contardo

    2012-01-01

      The CMS Upgrade Programme consists of four classes of projects: (a) Detector and Systems upgrades which are ongoing and largely (though not entirely) target LS1. (b) Full system upgrades for three projects that are preparing TDRs: Pixels, HCAL and L1 Trigger. The projects target completion by LS2. (c) Infrastructure consolidation and upgrades to improve operational robustness and to support the above projects. (d) Phase 2 replacement of the Tracker and major upgrades of the Trigger and Forward Detectors. For (a) and (c), detailed costing exists and is being integrated into a common reporting system. The schedule milestones for each project will be linked into the overall schedule planning for LS1. For the three TDR projects, the designs have progressed significantly since the Technical Proposal in 2010. Updated detailed cost estimates and schedules will be prepared with the TDRs to form the basis for tracking the projects through completion. To plan the upgrades and the supporting simulati...

  3. The Role of Nuclear Power in Slovak Republic; Safety Upgrading Program for WWER Reactors

    International Nuclear Information System (INIS)

    Toth, A.

    1998-01-01

    Implementation of Slovenske elektrarne Production Base Development Program, where all these safety upgrading projects of nuclear power sources at Slovak Republic are included will ensure first of all safety of nuclear power sources on internationally acceptable level, operational reliability of nuclear power units, balanced consumption and production of electric energy in Slovakia and decrease of long term ecological impact in according with international commitments of Slovakia

  4. ATR Technical Specification Upgrade Program

    International Nuclear Information System (INIS)

    McCracken, R.T.; Durney, J.L.; Freund, G.A.

    1990-01-01

    The Advanced Test Reactor (ATR) is a 250 MW, uranium-aluminum fueled test reactor which began full power operation in 1969. The initial operation was controlled by an Operating Limits document based on the original Safety Analysis Report. Additional safety bases were later developed to support Technical Specifications which were approved and implemented in 1977. The Technical Specifications which were initially developed with content and format specified in ANSI/ANS--15.1, ''The Development of Technical Specifications for Research Reactors.'' The safety basis documentation and the Technical Specifications have been updated as required to maintain them current with the ATR facility configuration. All revisions have been made with a content, format and style consistent with the original. A major, two-phase program to upgrade the content, format and style is in progress. This paper describes the first phase of this program

  5. Hanford Generic Interim Safety Basis

    International Nuclear Information System (INIS)

    Lavender, J.C.

    1994-01-01

    The purpose of this document is to identify WHC programs and requirements that are an integral part of the authorization basis for nuclear facilities that are generic to all WHC-managed facilities. The purpose of these programs is to implement the DOE Orders, as WHC becomes contractually obligated to implement them. The Hanford Generic ISB focuses on the institutional controls and safety requirements identified in DOE Order 5480.23, Nuclear Safety Analysis Reports

  6. Hanford Generic Interim Safety Basis

    Energy Technology Data Exchange (ETDEWEB)

    Lavender, J.C.

    1994-09-09

    The purpose of this document is to identify WHC programs and requirements that are an integral part of the authorization basis for nuclear facilities that are generic to all WHC-managed facilities. The purpose of these programs is to implement the DOE Orders, as WHC becomes contractually obligated to implement them. The Hanford Generic ISB focuses on the institutional controls and safety requirements identified in DOE Order 5480.23, Nuclear Safety Analysis Reports.

  7. Spent Nuclear Fuel (SNF) Project Safety Basis Implementation Strategy

    International Nuclear Information System (INIS)

    TRAWINSKI, B.J.

    2000-01-01

    The objective of the Safety Basis Implementation is to ensure that implementation of activities is accomplished in order to support readiness to move spent fuel from K West Basin. Activities may be performed directly by the Safety Basis Implementation Team or they may be performed by other organizations and tracked by the Team. This strategy will focus on five key elements, (1) Administration of Safety Basis Implementation (general items), (2) Implementing documents, (3) Implementing equipment (including verification of operability), (4) Training, (5) SNF Project Technical Requirements (STRS) database system

  8. Evolution of Safety Basis Documentation for the Fernald Site

    International Nuclear Information System (INIS)

    Brown, T.; Kohler, S.; Fisk, P.; Krach, F.; Klein, B.

    2004-01-01

    The objective of the Department of Energy's (DOE) Fernald Closure Project (FCP), in suburban Cincinnati, Ohio, is to safely complete the environmental restoration of the Fernald site by 2006. Over 200 out of 220 total structures, at this DOE plant site which processed uranium ore concentrates into high-purity uranium metal products, have been safely demolished, including eight of the nine major production plants. Documented Safety Analyses (DSAs) for these facilities have gone through a process of simplification, from individual operating Safety Analysis Reports (SARs) to a single site-wide Authorization Basis containing nuclear facility Bases for Interim Operations (BIOs) to individual project Auditable Safety Records (ASRs). The final stage in DSA simplification consists of project-specific Integrated Health and Safety Plans (I-HASPs) and Nuclear Health and Safety Plans (N-HASPs) that address all aspects of safety, from the worker in the field to the safety basis requirements preserving the facility/activity hazard categorization. This paper addresses the evolution of Safety Basis Documentation (SBD), as DSAs, from production through site closure

  9. Session 1 theme: Various forms of design basis knowledge and effects of its loss on Safety. Views from EDF

    International Nuclear Information System (INIS)

    Servière, Georges

    2013-01-01

    Design basis knowledge - What happens or may happen and corresponding required knowledge: • Unexpected events or failures of equipment; • Spare part issues (no longer availaible,…); • Change in applicable regulations / requirements; • Change of operating conditions; • Change of plant performances; • Evolution of external environment and conditions; • Events and accidents on other plants, worldwide; • New knowledge availaible; • Periodic safety reviews and upgrades; • Extension of plant operation life; • Decommissioning and dismantling; • Some of those you may choose not to do, but most of them have to be faced and need appropriate knowledge

  10. TPE upgrade for enhancing operational safety and improving in-vessel tritium inventory assessment in fusion nuclear environment

    Energy Technology Data Exchange (ETDEWEB)

    Shimada, M., E-mail: Masashi.Shimada@inl.gov [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Taylor, C.N.; Moore-McAteer, L.; Pawelko, R.J. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Kolasinski, R.D.; Buchenauer, D.A. [Sandia National Laboratories, Hydrogen and Materials Science Department, Livermore, CA 94550 (United States); Cadwallader, L.C.; Merrill, B.J. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States)

    2016-11-01

    The Tritium Plasma Experiment (TPE) is a unique high-flux linear plasma device that can handle beryllium, tritium, and neutron-irradiated plasma facing materials, and is the only existing device dedicated to evaluate in-vessel tritium inventory in the nuclear environment for fusion safety. The electrical upgrade were recently carried out to enhance operational safety and to improve plasma performance. New DC power supplies and a new control center enable remote plasma operations from outside of the contamination area for tritium, minimizing the possible exposure risk with tritium and beryllium and eliminating heat stress issue. In November 2015, the TPE successfully achieved first deuterium plasma via remote operation after a significant three-year upgrade. Simple linear scaling estimate showed that the TPE is expected to achieve Γ{sub i}{sup max} of >1.0 × 10{sup 23} m{sup −2} s{sup −1} and q{sub heat} of >1 MW m{sup −2} with new power supplies. This upgrade not only improves operational safety of the worker, but also enhances plasma performance to better simulate extreme plasma-material conditions expected in ITER, FNSF, and DEMO for improving in-vessel tritium inventory assessment in fusion nuclear environment.

  11. Safety upgrading activities against tsunami, earthquake, and severe accident at Hamaoka NPPs

    International Nuclear Information System (INIS)

    Watanabe, Tetsuya; Wakunaga, Takao; Ishida, Takahisa

    2013-01-01

    As the lessons learned by the Fukushima Daiichi NPPs accident, Chubu Electric Power carried out the Emergency Safety Measures at Hamaoka NPPs immediately, and announced the plan for tsunami countermeasures including the construction of 18m-height tsunami protection wall in July 2011. Furthermore, the company announced the additional severe accident and tsunami countermeasures, and etc. in December 2012 and in April 2013, such as the installation of Filtered Containment Venting System and increasing the height of the tsunami protection wall from 18m to 22m. In this paper, we present major safety upgrading activities against tsunami, earthquake and severe accident at Hamaoka NPPs. (author)

  12. Knowledge basis in safety culture for researchers and practitioners

    International Nuclear Information System (INIS)

    Vieira Neto, Antonio S.; Barroso, Antonio C.O.; Goncalves, Adriana

    2009-01-01

    This paper presents the main characteristics of the knowledge basis in safety culture which is being developed at the IPEN-CNEN/SP, one of the Brazilian nuclear institutes of research. The main objective of this basis is to organize the information about safety culture found in the literature and to make it available to researchers and practitioners. The first stage of the development of this basis is already finished being the subject of this work. (author)

  13. 340 waste handling facility interim safety basis

    Energy Technology Data Exchange (ETDEWEB)

    VAIL, T.S.

    1999-04-01

    This document presents an interim safety basis for the 340 Waste Handling Facility classifying the 340 Facility as a Hazard Category 3 facility. The hazard analysis quantifies the operating safety envelop for this facility and demonstrates that the facility can be operated without a significant threat to onsite or offsite people.

  14. 340 waste handling facility interim safety basis

    International Nuclear Information System (INIS)

    VAIL, T.S.

    1999-01-01

    This document presents an interim safety basis for the 340 Waste Handling Facility classifying the 340 Facility as a Hazard Category 3 facility. The hazard analysis quantifies the operating safety envelop for this facility and demonstrates that the facility can be operated without a significant threat to onsite or offsite people

  15. Upgradation of Apsara reactor

    International Nuclear Information System (INIS)

    Mammen, S.; Mukherjee, P.; Bhatnagar, A.; Sasidharan, K.; Raina, V.K.

    2009-01-01

    Apsara is a 1 MW swimming pool type research reactor using high enriched uranium as fuel with light water as coolant and moderator. The reactor is in operation for more than five decades and has been extensively used for basic research, radioisotope production, neutron radiography, detector testing, shielding experiments etc. In view of its long service period, it is planned to carry out refurbishment of the reactor to extend its useful life. During refurbishment, it is also planned to upgrade the reactor to a 2 MW reactor to improve its utilization and to upgrade the structure, system and components in line with the current safety standards. This paper gives a brief account of the design features and safety aspects of the upgraded Apsara reactor. (author)

  16. 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

  17. Criteria Document for B-plant's Surveillance and Maintenance Phase Safety Basis Document

    International Nuclear Information System (INIS)

    SCHWEHR, B.A.

    1999-01-01

    This document is required by the Project Hanford Managing Contractor (PHMC) procedure, HNF-PRO-705, Safety Basis Planning, Documentation, Review, and Approval. This document specifies the criteria that shall be in the B Plant surveillance and maintenance phase safety basis in order to obtain approval of the DOE-RL. This CD describes the criteria to be addressed in the S and M Phase safety basis for the deactivated Waste Fractionization Facility (B Plant) on the Hanford Site in Washington state. This criteria document describes: the document type and format that will be used for the S and M Phase safety basis, the requirements documents that will be invoked for the document development, the deactivated condition of the B Plant facility, and the scope of issues to be addressed in the S and M Phase safety basis document

  18. Safety upgrading of Novi Han Repository under IAEA TC Project BUL/4/005 achievements and future plans

    International Nuclear Information System (INIS)

    Stefanova, I.

    2003-01-01

    The report presents the safety upgrading of the Novi Han Repository under the IAEA TC Project BUL/4/005. The Project covers: identification of radionuclide inventory; characterisation of the disposal vaults; site characterisation; safety assessment; upgrading of the monitoring and radiation control; selection of treatment and conditioning processes and a conceptual design for a new waste processing and storage facility and other direct measures for safety improvement. The current inventory is identified and presented in the report. Schemes of the vault for solid wastes and vault for biological wastes are given, demonstrating reinforced concrete, stainless steel lining, and hydro insulation are presented. Several studies for safety assessment are made between 1997 and 2003. The operational safety assessment for disposal in existing facilities gives the annual risk for: spilling of waste package during upload (7.58.10 -9 ); spilling of waste package in transport accident (2.90.10 -9 ); fire scenario (3.50.10 - 1 3 ); radionuclide release due to flooding or earthquake (5.05.10 -4 ). The monitoring radiation control is upgraded according to the regulatory guidance and covered the site, restricted zone (1 km) and supervised zone (5 km). The types of analyses made are: Direct measurement of the dose rate -TLD; Direct measurements of the dose rate - portable surveillance monitors; In situ gamma spectrometry; Gamma spectrometry; Gross beta, gross alpha; Liquid scintillation spectrometry. The analyses show no transfer od radionuclides to the environment. The individual radiation control shows no evidence for specific radiation pathology. The operational radiation control service premises and transport vehicles. The following is measured: gamma dose rate; beta exposure; alpha exposure; neutron radiation; contamination level. Under the development is a detailed technical design supply of equipment for characterization of waste, including hot cell for control over high level

  19. Seismic safety review mission for the follow-up of the seismic upgrading of Kozloduy NPP (Units 1-4). Sofia, Bulgaria, 16-20 November 1992

    International Nuclear Information System (INIS)

    David, M.; Shibata, H.; Stevenson, J.D.; Godoy, A.; Gurpinar, A.

    1992-11-01

    A Seismic Safety Review Mission for the follow-up of the design and implementation of the seismic upgrading of Kozloduy NPP was performed in Sofia from 16-20 November 1992. This mission continued the second task of the follow-up activities of the design and implementation of the seismic upgrading (Phases 1 and 2), which is being carried out in Units 1 and 2 of the NPP. Thus the objectives of the mission was to assist the Bulgarian authorities in the technical evaluation of the design tasks defined for Phases 1 and 2 item HB of WANO 6 Month Programme, as follows: anchorage upgrades of low seismic capacity components; list of seismic safety related systems and components; detailed walkdown to assess seismic capacity of components and define priorities for the upgrading; determination of seismic structural capacity of pump house, diesel generator building and turbine building and design of required upgrades; liquefaction potential evaluation. Tabs

  20. Development of the irradiation facility SIBO INRA/Tangier, Morocco by upgrading cobalt-60 in a temporary pool and enhancing safety and control features

    Directory of Open Access Journals (Sweden)

    Mohammed Mouhib

    2017-12-01

    Full Text Available An automatic control system is one of the most important parts of an irradiation facility. The level of this control is always maintained to comply with safety procedures during routine work in this field. Also sometimes it is limited to the minimum level of regulation required due to economical aspects; some commercial systems are generally made by manufacturers of industrial facilities and considered affordable by irradiators. In some cases specific irradiation facilities tailor their control systems to their needs. For this kind of irradiator the control system can be developed and upgraded according to personal and industrial experiences. These upgrading procedures are also used by others to develop their systems. The objective of this paper is to share a local experience in upgrading security, safety systems and the use of cobalt-60 for the irradiator. It is a composite experiment at SIBO INRA/Tangier, Morocco and concerns the: (i upgrade of cobalt-60 in a temporary pool in the SIBO irradiator in Tangier. This operation was conducted in collaboration with the International Atomic Energy Agency (IAEA and was a success story of 2014 according to the general conference of IAEA; (ii safety and technical upgrade of the system in the SIBO irradiator made in collaboration with IAEA; (iii installation and upgrade of the security system in accordance with the Global Threat Reduction Programme (GTRP to reduce the threat of a Radiological Dispersal Device (RDD in collaboration with The United States Department of Energy’s National Nuclear Security Administration (NNSA.

  1. Risk management considerations for seismic upgrading of an older facility for short-term residue stabilization

    International Nuclear Information System (INIS)

    Additon, S.L.; Peregoy, W.L.; Foppe, T.L.

    1999-01-01

    Building 707 and its addition, Building 707A, were selected, after the production mission of Rocky Flats was terminated a few years ago, to stabilize many of the plutonium residues remaining at the site by 2002. The facility had undergone substantial safety improvements to its safety systems and conduct of operations for resumption of plutonium operations in the early 1990s and appeared ideally suited for this new mission to support accelerated Site closure. During development of a new authorization basis, a seismic evaluation was performed. This evaluation addressed an unanalyzed expansion joint and suspect connection details for the precast concrete tilt-up construction and concluded that the seismic capacity of the facility is less than half of that determined by previous analysis. Further, potential seismic interaction was identified between a collapsing Building 707 and the seismically upgraded Building 707A, possibly causing the partial collapse of the latter. Both the operating contractor and the Department of Energy sought a sound technical basis for deciding how to proceed. This paper addresses the risks of the as-is facility and possible benefits of upgrades to support a decision on whether to upgrade the seismic capacity of Building 707, accept the risk of the as-is facility for its short remaining mission, or relocate critical stabilization missions. The paper also addresses the Department of Energy's policy on natural phenomena

  2. Solid waste burial grounds interim safety analysis

    International Nuclear Information System (INIS)

    Saito, G.H.

    1994-01-01

    This Interim Safety Analysis document supports the authorization basis for the interim operation and restrictions on interim operations for the near-surface land disposal of solid waste in the Solid Waste Burial Grounds. The Solid Waste Burial Grounds Interim Safety Basis supports the upgrade progress for the safety analysis report and the technical safety requirements for the operations in the Solid Waste Burial Grounds. Accident safety analysis scenarios have been analyzed based on the significant events identified in the preliminary hazards analysis. The interim safety analysis provides an evaluation of the operations in the Solid Waste Burial Grounds to determine if the radiological and hazardous material exposures will be acceptable from an overall health and safety standpoint to the worker, the onsite personnel, the public, and the environment

  3. Solid waste burial grounds interim safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Saito, G.H.

    1994-10-01

    This Interim Safety Analysis document supports the authorization basis for the interim operation and restrictions on interim operations for the near-surface land disposal of solid waste in the Solid Waste Burial Grounds. The Solid Waste Burial Grounds Interim Safety Basis supports the upgrade progress for the safety analysis report and the technical safety requirements for the operations in the Solid Waste Burial Grounds. Accident safety analysis scenarios have been analyzed based on the significant events identified in the preliminary hazards analysis. The interim safety analysis provides an evaluation of the operations in the Solid Waste Burial Grounds to determine if the radiological and hazardous material exposures will be acceptable from an overall health and safety standpoint to the worker, the onsite personnel, the public, and the environment.

  4. Interim safety basis compliance matrix for Trenches 31 and 34

    International Nuclear Information System (INIS)

    Ames, R.R.

    1994-01-01

    The tables provided in this document identify the specific requirements and basis for the administrative controls established in the Westinghouse Hanford Company (WHC) Solid Waste Burial Ground (SWBG) Interim Safety Basis (ISB) for operation of the Project W-025, Mixed Waste Lined Landfill (Trenches 31 and 34). The tables document the necessary controls and implementing procedures to ensure compliance with the requirements of the ISB. These requirements provide a basis for future Unreviewed Safety Questions (USQ) screening of applicable procedure changes, proposed physical modifications, tests, experiments, and occurrences. Table 1 provides the SWBG interim Operational Safety Requirements administrative controls matrix. The specific assumptions and commitments used in the safety analysis documents applicable to disposal of mixed wastes in Trenches 31 and 34 are provided in Table 2. Table 3 is provided to document the potential engineered and administrative mitigating features identified in the Preliminary Hazard Analysis (PHA) for disposal of mixed waste

  5. Status of research reactors in China. Their utilization and safety upgrading

    International Nuclear Information System (INIS)

    Xu Hanming; Jin Huajin

    2000-01-01

    The main research reactors in China basically consist of several old reactors including HWRR, HFETR, SPR, MJTR and MNSR. Except the last one, all the other reactors operate at a high power density and represent themselves as main tools in China for engineering testing, radioactive isotope production, and neutron scattering research. The research and production activities by these reactors are briefed. Main equipment and research topics for neutron scattering are described. The production of radioisotope is summarized. Safety upgrading activities in recent years taken by these old reactors are described, which make the safety feature of each reactor significantly improved and on the whole more close to (even not completely consistent) with the targets set by the modern safety regulation. Since a new multi-purpose research reactor CARR is expected available around the year of 2005, a schedule about the construction of new reactor, reforming or decommissioning of old reactors and smoothly transition of research and production activities from old to new reactor during the coming years has been under careful planning. A suggestion of potential international cooperation items has been preliminarily given. (author)

  6. Upgrading of TREAT experimental capabilities

    International Nuclear Information System (INIS)

    Dickerman, C.E.; Rose, D.; Bhattacharyya, S.K.

    1982-01-01

    The TREAT facility at the Argonne National Laboratory site in the Idaho National Engineering Laboratory is being upgraded to provide capabilities for fast-reactor-safety transient experiments not possible at any other experimental facility. Principal TREAT Upgrade (TU) goal is provision for 37-pin size experiments on energetics of core-disruptive accidents (CDA) in fast breeder reactor cores with moderate sodium void coefficients. this goal requires a significant enhancement of the capabilities of the TREAT facility, specifically including reactor control, hardened neutron spectrum incident on the test sample, and enlarged building. The upgraded facility will retain the capability for small-size experiments of the types currently being performed in TREAT. Reactor building and crane upgrading have been completed. TU schedules call for the components of the upgraded reactor system to be finished in 1984, including upgraded TREAT fuel and control system, and expanded coverage by the hodoscope fuel-motion diagnostics system

  7. Seismic assessment and upgrading of nuclear power plants in Eastern Europe

    Energy Technology Data Exchange (ETDEWEB)

    Katona, T; Kostov, M

    1997-03-01

    The basic findings of the seismic re-qualification programmes going on recently at all VVER plants in Eastern Europe can be summarised. The problems of the seismic safety have to be solved taking into account the general concept of the nuclear safety enhancement of the units. There are cases where the system improvements lead to better and more effective solution of the problem than the structural upgrading. The equipment and piping of the primary system have sufficient capacity. The viscous dampers are considered usually for the upgrading. The equipment anchorage especially the electrical and I and C equipment anchorage have to be upgraded. There are general consideration for replacement of the hydraulic snubbers by viscous dampers in the primary circuit of the VVER 440/V230. The considerations are not only because of the better seismic behaviour but mainly because of the better operational performance. There is relatively good seismic instrumentation at the plants considered. The definition of the scram level of the units not designed for an OBE is an essential problem. More effort needed for the definition of this level on the basis of re-evaluation experience of the plant equipment and after the proper definition of post-earthquake activities. The seismic re-evaluation and re-qualification of the VVER units is a general safety issue in Easter European countries. This rather complex problem can be solved adopting the experience, methods and requirements of western countries and taking into account the design features of the VVER units as well as the as built and as it is conditions. (J.P.N.)

  8. Seismic assessment and upgrading of nuclear power plants in Eastern Europe

    International Nuclear Information System (INIS)

    Katona, T.; Kostov, M.

    1997-01-01

    The basic findings of the seismic re-qualification programmes going on recently at all VVER plants in Eastern Europe can be summarised. The problems of the seismic safety have to be solved taking into account the general concept of the nuclear safety enhancement of the units. There are cases where the system improvements lead to better and more effective solution of the problem than the structural upgrading. The equipment and piping of the primary system have sufficient capacity. The viscous dampers are considered usually for the upgrading. The equipment anchorage especially the electrical and I and C equipment anchorage have to be upgraded. There are general consideration for replacement of the hydraulic snubbers by viscous dampers in the primary circuit of the VVER 440/V230. The considerations are not only because of the better seismic behaviour but mainly because of the better operational performance. There is relatively good seismic instrumentation at the plants considered. The definition of the scram level of the units not designed for an OBE is an essential problem. More effort needed for the definition of this level on the basis of re-evaluation experience of the plant equipment and after the proper definition of post-earthquake activities. The seismic re-evaluation and re-qualification of the VVER units is a general safety issue in Easter European countries. This rather complex problem can be solved adopting the experience, methods and requirements of western countries and taking into account the design features of the VVER units as well as the as built and as it is conditions. (J.P.N.)

  9. The role of IAEA in the seismic assessment and upgrading of existing NPPs. Seismic safety of nuclear power plants in Eastern Europe

    Energy Technology Data Exchange (ETDEWEB)

    Guerpinar, A; Godoy, A [International Atomic Energy Agency, Vienna (IAEA). Div. of Nuclear Installation Safety

    1997-03-01

    This paper summarizes the work performed by the International Atomic Energy Agency in the areas of safety reviews and applied research in support of programmes for the assessment and enhancement of seismic safety in Eastern Europe and in particular WWER type nuclear power plants during the past seven years. Three major topics are discussed; engineering safety review services in relation to external events, technical guidelines for the assessment and upgrading of WWER type nuclear power plants, and the Coordinated Research Programme on `Benchmark study for the seismic analysis and testing of WWER type nuclear power plants`. These topics are summarized in a way to provide an overview of the past and present safety situation in selected WWER type plants which are all located in Eastern European countries. Main conclusion of the paper is that although there is now a thorough understanding of the seismic safety issues in these operating nuclear power plants, the implementation of seismic upgrades to structures, systems and components are lagging behind, particularly for those cases in which the re-evaluation indicated the necessity to strengthen the safety related structures or install new safety systems. (author)

  10. The role of IAEA in the seismic assessment and upgrading of existing NPPs. Seismic safety of nuclear power plants in Eastern Europe

    International Nuclear Information System (INIS)

    Guerpinar, A.; Godoy, A.; . Div. of Nuclear Installation Safety)

    1997-01-01

    This paper summarizes the work performed by the International Atomic Energy Agency in the areas of safety reviews and applied research in support of programmes for the assessment and enhancement of seismic safety in Eastern Europe and in particular WWER type nuclear power plants during the past seven years. Three major topics are discussed; engineering safety review services in relation to external events, technical guidelines for the assessment and upgrading of WWER type nuclear power plants, and the Coordinated Research Programme on 'Benchmark study for the seismic analysis and testing of WWER type nuclear power plants'. These topics are summarized in a way to provide an overview of the past and present safety situation in selected WWER type plants which are all located in Eastern European countries. Main conclusion of the paper is that although there is now a thorough understanding of the seismic safety issues in these operating nuclear power plants, the implementation of seismic upgrades to structures, systems and components are lagging behind, particularly for those cases in which the re-evaluation indicated the necessity to strengthen the safety related structures or install new safety systems. (author)

  11. Experience of upgrading existing Russian designed nuclear plants

    International Nuclear Information System (INIS)

    Yanev, P.I.; Facer, R.I.

    1993-01-01

    From the reviewed experiences of upgrading existing Russian designed nuclear plants both of WWER and RBMK type the conclusions drawn are as follows. For the countries operating Russian designed plants it is necessary to adopt a pragmatic approach where all changes must be demonstrated to improve the safety of the plant and safety must be demonstrably improving. Care must be taken to avoid the pitfalls of excessive regulatory demands which are not satisfied and the development of an attitude of disregarding requirements on the basis that they are not enforced. For the lending countries and organizations, it is necessary to ensure that assistance is given to the operating organizations so that the most effective use of funds can be achieved. The experience in the West is that over-regulation and excessive expenditure do not necessarily lead to improved safety. They can lead to significant waste of resources. The use of western technology is recommended but where it is necessary and where it provides the greatest benefit

  12. Upgrade Uranium Recovery Project No. 34110: final safety analysis report

    International Nuclear Information System (INIS)

    1981-09-01

    The accident analysis of the upgrade uranium recovery system indicated three potential hazards: (1) criticality, (2) toxic fumes from nitric acid solutions, and (3) release of toxic uranyl nitrate solutions. Any of these are capable of causing the death of one or more employees; therefore, they form the basis for the residual risks identified below. The analysis found no hazardous energies or substances capable of causing irreversible injury to, or the death of, any members of the public. The following residual risks will be controlled administratively by procedural constraints: An operator or maintenance error will cause 235 U to be transferred into an unsafe container and cause a criticality. An operator or maintenance error will cause containers of 235 U bearing material to be improperly spaced and cause a criticality. Extensive corrosion will cause a hole to form in a calciner tube, the corrosion will go undetected, and a criticality will result, and a loss of system and/or building solution containment will occur concurrent with a drain being open resulting in a criticality and/or release of toxic material. Additional residual risks that have a small probability are that an earthquake or tornado will affect the building, alter the system geometry, and initiate a criticality; that the compressed-gas (nitrogen) cylinder valve will be sheared off, become airborne, and alter the system geometry; and that loss of system and/or building solution containment may occur concurrently with fire sprinkler system actuation causing a criticality and/or release of toxic material. The following residual risks will be addressed in the Safety Study of the existing X-705 Building: that a spill of raffinate highly contaminated with 99 Tc will occur due to operator error or incorrect lab analysis and that a gaseous or liquid effluent release of small amounts of transuranic elements will occur

  13. Constructing an I and C Upgrade Architecture for Korea Standard Nuclear Power Plants

    International Nuclear Information System (INIS)

    Suh, Yong Suk; Hur, Seop; Kim, Dong Hoon; Sung, Chan Ho; Kang, Hyun Tai; Lee, Jae Ki; Cho, Chang Hwan

    2008-01-01

    This paper presents three architectures of the KSNP I and C upgrade. The architectures are constructed with an adoption of PLCs and DCS technology and 3-phase upgrade strategy. The 3-phase upgrade strategy is established to ensure the safety of the upgrade. Based on the architecture, the cabinet configuration is being constructed. From the configuration, it is expected to figure out how to optimize the layout of the cabinets. It is required to study the performance and safety design requirements of the upgrade further

  14. Bohunice V1 NPP upgrading programme

    International Nuclear Information System (INIS)

    Kerak, J.

    2001-01-01

    The paper describes whole process of Bohunice V1 NPP nuclear safety and operational reliability level increase which has been performed since units commissioning (1. unit in 1978, 2. unit in 1980), continued Small Reconstruction (1991 -1993) and finished Gradual Upgrading(1994 -2000). The main purpose is to last stage -Gradual upgrading of Bohunice V1 NPP. (author)

  15. Instrumentation and control doctrine for VVER upgrading

    International Nuclear Information System (INIS)

    Pernette, R.; Cucciniello, J.; Trebot, R.

    1997-01-01

    The main results are reported of analyses carried out on the WWER-400 and WWER-1000 series, and the reasons are summed up underlying the decision to establish a policy of WWER instrumentation and control upgrading. It is shown that such a policy forms a key stage in any upgrading project and contributes to the improvement of reactor safety at several levels: it helps to optimize the integration of safety requirements; it makes it easier to achieve a high level of automation and a better man-machine interaction; it should raise the culture of reactor safety. Four annexes demonstrate the factual fundamentals and structure of such a policy. (A.K.)

  16. Nuclear safety requirements for upgrading the National Repository for Radioactive Wastes-Baita Bihor

    International Nuclear Information System (INIS)

    Vladescu, Gabriela; Necula, Daniela

    2000-01-01

    The upgrading project of National Repository for Radioactive Wastes-Baita Bihor is based on the integrated concept of nuclear safety. Its ingredients are the following: A. The principles of nuclear safety regarding the management of radioactive wastes and radioprotection; B. Safety objectives for final disposal of low- and intermediate-level radioactive wastes; C. Safety criteria for final disposal of low- and intermediate-level radioactive wastes; D. Assessment of safety criteria fulfillment for final disposal of low- and intermediate-level radioactive wastes. Concerning the nuclear safety in radioactive waste management the following issues are considered: population health protection, preventing transfrontier contamination, future generation radiation protection, national legislation, control of radioactive waste production, interplay between radioactive waste production and management, radioactive waste repository safety. The safety criteria of final disposal of low- and intermediate-level radioactive wastes are discussed by taking into account the geological and hydrogeological configuration, the physico-chemical and geochemical characteristics, the tectonics and seismicity conditions, extreme climatic potential events at the mine location. Concerning the requirements upon the repository, the following aspects are analyzed: the impact on environment, the safety system reliability, the criticality control, the filling composition to prevent radioactive leakage, the repository final sealing, the surveillance. Concerning the radioactive waste, specific criteria taken into account are the radionuclide content, the chemical composition and stability, waste material endurance to heat and radiation. The waste packaging criteria discussed are the mechanical endurance, materials toughness and types as related to deterioration caused by handling, transportation, storing or accidents. Fulfillment of safety criteria is assessed by scenarios analyses and analyses of

  17. Industry perspective on digital upgrades: A utility perspective on digital upgrades

    International Nuclear Information System (INIS)

    Blauw, R.J.

    1994-01-01

    Nuclear utilities face the need to upgrade aging and obsolete safety related and other critical equipment. This is the result of operation and maintenance concerns for reliability and maintainability. Digital technology is an option for these upgrades. A number of utilities have attempted exercising the digital option. The regulatory licensing results have been inconsistent and have raised a variety of issues. These issues and the subsequent licensing uncertainties have caused some utilities to temporarily drop digital technology as an upgrade option. Resolution of these issues and the need for regulatory stability is driving the development of industry standards and guidelines. These will provide guidance to support consistent design and implementation of digital upgrades. Successful completion of these documents is necessary for renewed consideration of the use of digital technology. This paper will present a utility perspective on how project management, configuration control, and a rigorous design process can serve to address the present regulatory issues. These issues include commercial grade dedication, reliability, electromagnetic interference, and failure and error management. This perspective is consistent with the standards and guidelines development effort

  18. Upgrading of radon's type near surface repository in Latvia

    International Nuclear Information System (INIS)

    Abramenkovs, A.

    2006-01-01

    In 1959, the Soviet government decided to construct the near surface radioactive wastes repository 'Radons' near the Baldone city. It was put in operation in 1962. The changes in the development of the repository were induced by the necessarily to upgrade it for disposal of radioactive wastes from the decommissioning of the Salaspils Research Reactor (SRR). The safety assessment of repository was performed during 2000-2001 under the PHARE project for necessary upgrades of repository. The outline design for new vaults and interim storage for long lived radioactive wastes was elaborated during 2003-2004 years. The Environmental Impact Assessment (EIA) for upgrade of Baldone repository was performed during 2004-2005 years. It was found, that additional efforts must be devoted for solution of social aspects o for successful operation and upgrade of repository. It was shown by EIA, that the local population has a negative opinion against the upgrade of repository in Latvia. The main recommendations for upgrades were connected with increasing the safety of repository, increasing of PR activities for education of society and developing of compensation mechanism for local municipality. (author)

  19. Upgrading of regulatory activities in Belarus

    International Nuclear Information System (INIS)

    Rozdyalovskaya, L.F.; Shabanov, V.V.

    1998-01-01

    Upgrading of the National regulatory regime in the Republic of Belarus started in fact in 1992 after the Interregional Nuclear and Radiation Safety Inspectorate had been established in the Gostekhnadzor of Belarus. In this connection, the Gostekhnadzor was transformed into the Gospromatomnadzor - the State Committee for Supervision of Industrial and Nuclear Safety. In 1993, by special decrees issued by the Council of Ministers the Gospromatomnadzor was designated a National Competent Body responsible for nuclear materials. Now the Committee is part of the Ministry for Emergencies. In the Committee, the issues of nuclear and radiation safety are dealt with by 6 departments among which the Department for Nuclear and Radiation Safety Regulation and the Interregional Nuclear and Radiation Safety inspectorate play major role. The created regulatory structure makes it possible to fully perform the control of situation at nuclear- and radiation-hazardous facilities and take adequate measures aimed at strengthening their safety. The priority directions of regulatory work by the Promatomnadzor include development and revision of regulations governing radiation and nuclear safety and upgrading of the training procedure to enhance the radiation safety and technical knowledge of the Promatomnadzor staff. (author)

  20. Safety basis for the 241-AN-107 mixer pump installation and caustic addition

    International Nuclear Information System (INIS)

    Van Vleet, R.J.

    1994-01-01

    This safety Basis was prepared to determine whether or not the proposed activities of installing a 76 HP jet mixer pump and the addition of approximately 50,000 gallons of 19 M (50:50 wt %) aqueous caustic are within the safety envelope as described by Tank Farms (chapter six of WHC-SD-WM-ISB-001, Rev. 0). The safety basis covers the components, structures and systems for the caustic addition and mixer pump installation. These include: installation of the mixer pump and monitoring equipment; operation of the mixer pump, process monitoring equipment and caustic addition; the pump stand, caustic addition skid, the electrical skid, the video camera system and the two densitometers. Also covered is the removal and decontamination of the mixer pump and process monitoring system. Authority for this safety basis is WHC-IP-0842 (Waste Tank Administration). Section 15.9, Rev. 2 (Unreviewed Safety Questions) of WHC-IP-0842 requires that an evaluation be performed for all physical modifications

  1. Combination of Eastern and Western technology in VVER 1000 NPP upgrade

    International Nuclear Information System (INIS)

    Ubra, O.

    1997-01-01

    An extensive modernization program is presented for upgrading the two WWER-type units of nuclear power plant Temelin to meet the latest international safety standards. The following innovations have been implemented: modernization and upgrading of the safety and control systems, new fuel design and modification of the reactor core, new diagnostic system, innovations of some components and subsystems of the primary and secondary systems, design and construction of a full scope simulator, the improvement of safety documentation, development of the Probabilistic Safety Assessment programme. (M.D.)

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

    International Nuclear Information System (INIS)

    Gibson, K.D.

    1995-09-01

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

  3. Central waste complex interim safety basis

    International Nuclear Information System (INIS)

    Cain, F.G.

    1995-01-01

    This interim safety basis provides the necessary information to conclude that hazards at the Central Waste Complex are controlled and that current and planned activities at the CWC can be conducted safely. CWC is a multi-facility complex within the Solid Waste Management Complex that receives and stores most of the solid wastes generated and received at the Hanford Site. The solid wastes that will be handled at CWC include both currently stored and newly generated low-level waste, low-level mixed waste, contact-handled transuranic, and contact-handled TRU mixed waste

  4. Improvements for real-time magnetic equilibrium reconstruction on ASDEX Upgrade

    International Nuclear Information System (INIS)

    Giannone, L.; Fischer, R.; McCarthy, P.J.; Odstrcil, T.; Zammuto, I.; Bock, A.; Conway, G.; Fuchs, J.C.; Gude, A.; Igochine, V.; Kallenbach, A.; Lackner, K.; Maraschek, M.; Rapson, C.; Ruan, Q.; Schuhbeck, K.H.; Suttrop, W.; Wenzel, L.

    2015-01-01

    Highlights: • Spline basis current functions with second-order linear regularisation. • Perturbations of magnetic probe measurements due to ferromagnetic tiles on the inner wall and from oscillations in the fast position coil current are corrected. • A constraint of the safety factor on the magnetic axis is introduced. Soft X-ray tomography is used to assess the quality of the real-time magnetic equilibrium reconstruction. • External loop voltage measurements and magnetic probe pairs inside and outside the vessel wall were used to measure the vacuum vessel wall resistivity. - Abstract: Real-time magnetic equilibria are needed for NTM stabilization and disruption avoidance experiments on ASDEX Upgrade. Five improvements to real-time magnetic equilibrium reconstruction on ASDEX Upgrade have been investigated. The aim is to include as many features of the offline magnetic equilibrium reconstruction code in the real-time equilibrium reconstruction code. Firstly, spline current density basis functions with regularization are used in the offline equilibrium reconstruction code, CLISTE [1]. It is now possible to have the same number of spline basis functions in the real-time code. Secondly, in the presence of edge localized modes, (ELM's), it is found to be necessary to include the low pass filter effect of the vacuum vessel on the fast position control coil currents to correctly compensate the magnetic probes for current oscillations in these coils. Thirdly, the introduction of ferromagnetic tiles in ASDEX Upgrade means that a real-time algorithm for including the perturbations of the magnetic equilibrium generated by these tiles is required. A methodology based on tile surface currents is described. Fourthly, during current ramps it was seen that the difference between fitted and measured magnetic measurements in the equilibrium reconstruction were larger than in the constant current phase. External loop voltage measurements and magnetic probe pairs inside and

  5. Improvements for real-time magnetic equilibrium reconstruction on ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Giannone, L.; Fischer, R. [Max Planck Institute for Plasma Physics, 85748 Garching (Germany); McCarthy, P.J. [Department of Physics, University College Cork, Cork (Ireland); Odstrcil, T.; Zammuto, I.; Bock, A.; Conway, G.; Fuchs, J.C.; Gude, A.; Igochine, V.; Kallenbach, A.; Lackner, K.; Maraschek, M.; Rapson, C. [Max Planck Institute for Plasma Physics, 85748 Garching (Germany); Ruan, Q. [National Instruments, Austin, TX 78759-3504 (United States); Schuhbeck, K.H.; Suttrop, W. [Max Planck Institute for Plasma Physics, 85748 Garching (Germany); Wenzel, L. [National Instruments, Austin, TX 78759-3504 (United States)

    2015-11-15

    Highlights: • Spline basis current functions with second-order linear regularisation. • Perturbations of magnetic probe measurements due to ferromagnetic tiles on the inner wall and from oscillations in the fast position coil current are corrected. • A constraint of the safety factor on the magnetic axis is introduced. Soft X-ray tomography is used to assess the quality of the real-time magnetic equilibrium reconstruction. • External loop voltage measurements and magnetic probe pairs inside and outside the vessel wall were used to measure the vacuum vessel wall resistivity. - Abstract: Real-time magnetic equilibria are needed for NTM stabilization and disruption avoidance experiments on ASDEX Upgrade. Five improvements to real-time magnetic equilibrium reconstruction on ASDEX Upgrade have been investigated. The aim is to include as many features of the offline magnetic equilibrium reconstruction code in the real-time equilibrium reconstruction code. Firstly, spline current density basis functions with regularization are used in the offline equilibrium reconstruction code, CLISTE [1]. It is now possible to have the same number of spline basis functions in the real-time code. Secondly, in the presence of edge localized modes, (ELM's), it is found to be necessary to include the low pass filter effect of the vacuum vessel on the fast position control coil currents to correctly compensate the magnetic probes for current oscillations in these coils. Thirdly, the introduction of ferromagnetic tiles in ASDEX Upgrade means that a real-time algorithm for including the perturbations of the magnetic equilibrium generated by these tiles is required. A methodology based on tile surface currents is described. Fourthly, during current ramps it was seen that the difference between fitted and measured magnetic measurements in the equilibrium reconstruction were larger than in the constant current phase. External loop voltage measurements and magnetic probe pairs inside

  6. Risk-informed decision making during Bohunice NPP safety upgrading

    International Nuclear Information System (INIS)

    Lipar, M.; Muzikova, E.; Kubanyi, J.

    2001-01-01

    The paper summarizes some facts of risk-informed regulation developments within UJD regulatory environment. Based on national as well as international operating experience and indications resulted from PSA, Nuclear Regulatory Authority of the Slovak Republic (UJD) since its constituting in 1993 has devoted an effort to use PSA technology to support the regulatory policy in Slovakia. The PSA is considered a complement, not a substitute, to the deterministic approach. Suchlike integrated approach is used in decision making processes and the final decision on scope and priorities is based on it. The paper outlines risk insights used in the decision making process concerning Bohunice NPP safety upgrading and focuses on the role of PSA results in Gradual Reconstruction of Bohunice VI NPP. Besides, two other examples of the PSA results application to the decision making process are provided: the assessment of proposal of modifications to the main power supply diagram (incorporation of generator switches) and the assessment of licensee request for motor generator AOT (Allowable Outage Time) extension. As an example of improving support of Bohunice V-2 risk-informed operations, concept of AOT calculations and Bohunice V-2 Risk Monitor Project are briefly described. (author)

  7. Authorization basis status report (miscellaneous TWRS facilities, tanks and components)

    Energy Technology Data Exchange (ETDEWEB)

    Stickney, R.G.

    1998-04-29

    This report presents the results of a systematic evaluation conducted to identify miscellaneous TWRS facilities, tanks and components with potential needed authorization basis upgrades. It provides the Authorization Basis upgrade plan for those miscellaneous TWRS facilities, tanks and components identified.

  8. Authorization basis status report (miscellaneous TWRS facilities, tanks and components)

    International Nuclear Information System (INIS)

    Stickney, R.G.

    1998-01-01

    This report presents the results of a systematic evaluation conducted to identify miscellaneous TWRS facilities, tanks and components with potential needed authorization basis upgrades. It provides the Authorization Basis upgrade plan for those miscellaneous TWRS facilities, tanks and components identified

  9. A defence in depth approach to safety assessment of existing nuclear power plant

    International Nuclear Information System (INIS)

    Butcher, P.; Holloway, N.J.

    1998-01-01

    The safety assessment of plant built to earlier standards requires an approach to prioritisation of upgrades that is based on sound engineering and safety principles. The principles of defence in depth are universally accepted and can form the basis of a prioritisation scheme for safety issues, and hence for the upgrading required to address them. The described scheme includes criteria for acceptability and issue prioritisation that are based on the number of lines of defence and the consequences of their failure. They are thus equivalent in concept to risk criteria, but are based on deterministic principles. This scheme has been applied successfully to the RBMK plant at Ignalina in Lithuania, for which a Western-style Safety Analysis Report has recently been produced and reviewed by joint Western and Eastern teams. An extended Safety Improvement Programme (SIP2) has been developed and agreed, based on prioritisations from the defence in depth assessment. (author)

  10. Upgrade of Instrumentation for Purdue Reactor PUR-1

    International Nuclear Information System (INIS)

    Revankar, S.T.; Merritt, E.; Bean, R.

    2000-01-01

    The major objective of this program was to upgrade and replace instruments and equipment that significantly improve the performance, control and operational capability of the Purdue University nuclear reactor (PUR-1). Under this major objective two projects on instrument upgrade were implemented. The first one was to convert the vacuum tube control and safety amplifiers (CSA) to solid state electronics, and the other was to upgrade the electrical and electronic shielding. This report is the annual report and gives the efforts and progress achieved on these two projects from July 1999 to June 2000

  11. Safety improvements made at the Loviisa nuclear power plant to reduce fire risks originating from the turbine generators

    International Nuclear Information System (INIS)

    Virolainen, T.; Marttila, J.; Aulamo, H.

    1998-01-01

    Comprehensive upgrading measures have been completed for the Loviisa Nuclear Power Plant (modified VVER440/V213). These were carried out from the start of the design phase and during operation to ensure safe plant shutdown in the event of a large turbine generator oil fire. These modifications were made mainly on a deterministic basis according to specific risk studies and fire analyses. As part of the probabilistic safety assessment, a fire risk analysis was made that confirmed the importance of these upgrading measures. In fact, they should be considered as design basis modifications for all VVER440 plants. (author)

  12. Preliminary studies for the upgrading of the ININ TRIGA MARK III reactor

    International Nuclear Information System (INIS)

    Aguilar, J.V.

    1984-01-01

    An increase on the demand of services has led to the -consideration of reactor upgrading, for which a project has been proposed. This project includes the modification of the reactor core, using FLIP type (Erbium loaded, 70% U-235 enriched) fuel elements and the upgrading of the control - console which implies a careful analysis of the Safety -considerations involved in the system modification. In this paper, a brief description is given of: The modifications intended; Pre-operational tests for the system components to be modified; The safety report needed for the licensing of the upgraded reactor; The project management and evaluation

  13. Analysis and upgrade of instrumentation and control systems for the modernization of research reactors

    International Nuclear Information System (INIS)

    1988-01-01

    This document provides assistance in the review and planning process for the upgrade of instrumentation and control systems (I and C systems) and related safety features of the reactor protection system for research reactors. In the interest of safety a need was realized to evaluate the performance of outdated I and C systems. An advisory group was assembled to develop guidelines and to provide recommendations for the upgrade of I and C systems. The recommendations on I and C systems upgrade contained in this document were developed by the advisory group using as guidelines the established safety criteria and operating standards for research reactors. 24 refs

  14. Replacement cross-site transfer system project W-058 safety class upgrade summary report

    International Nuclear Information System (INIS)

    Schlosser, R.L.

    1998-01-01

    This report evaluates the design of the replacement cross-site transfer system structures, systems, and components for safety related applications as defined in the Tank Waste Remediation Systems Basis for Interim Operations

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

    Energy Technology Data Exchange (ETDEWEB)

    David Petti; Jim Kinsey; Dave Alberstein

    2014-01-01

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

  16. Safety evaluation for instrumentation and control system upgrading project of Malaysian TRIGA MARK II PUSPATI Research reactor

    International Nuclear Information System (INIS)

    Ridha Roslan; Nik Mohd Faiz Khairuddin

    2013-01-01

    Full-text: Malaysian TRIGA MARK II research reactor has been in safe operation since its first criticality in 1982. The reactor is licensed to be operated by Malaysian Nuclear Agency to perform training and research development related activities. Due to its extensive operation since last three decades, the option of modifications for safety and safety-related item and component become a necessary to replace the outdated equipment to a stat-of-art, reliable technologies. This paper will present the current regulatory activities performed by Atomic Energy Licensing Board (AELB) to ensure the upgrading of analogue to digital instrumentation and control system is implemented in safe manner. The review activity includes documentation review, manufacturer quality audit and on-site inspection for commissioning. The review performed by AELB is based on The International Atomic Energy Agency (IAEA) Safety Requirements NS-R-4, entitled Safety of Research Reactors. During this endeavour, AELB seeks technical cooperation from Korea Institute of Nuclear Safety (KINS), the nuclear experts organization of the country of origin of the instrumentation and control technology. The regulatory activity is still on-going and is expected to be completed by issuance of Authorization for Restart on December 2013. (author)

  17. Upgrading Atucha 1 nuclear power plant. Regulatory perspective

    International Nuclear Information System (INIS)

    Caruso, G.

    1998-01-01

    Atucha 1 nuclear power plant has unique design and its commercial operation started in 1974. The upgrading decisions, the basis for an upgrading program and its status of implementation are presented. Regulatory decisions derived from the performance-based approach have the advantage that they enable balancing of the overall plant risk and identifying at different plant levels the areas where improvements are necessary. (author)

  18. DEVELOPING SAFETY INDICATORS ON THE BASIS OF THE ICAO RECOMMENDATIONS

    Directory of Open Access Journals (Sweden)

    V. D. Sharov

    2014-01-01

    Full Text Available The article offers direct use of the recommendations of SMM ICAO Doc.9859, 3rd ed. 2013, for calculation the target and alert levels of safety indicators. Examples of calculation based on data of 2011 and monitoring of the current indicators during 2012 are presented. Safety indicators for airlines in terms of “numbers of incidents per 1000 flight hours” could be calculated on the basis of the state values through the «coefficient of conformity».

  19. Original seismic and similar severe external loading design basis for WWER type nuclear power plants in Czech and Slovak Republics and actual issues of their upgrading

    International Nuclear Information System (INIS)

    Masopust, R.

    1993-01-01

    The WWER type NPPs located in Czech and Slovak republics have many seismic vulnerabilities similar to those recognized in many of the US NPPs prior to late seventies. They are mostly caused by underestimation of these problems in the design phases, sometimes due to inadequate performance and poor quality of works and some incompatibilities between the original Russian design and current international design bases and safety requirements. It is believed that the structures and equipment of these NPPs can be seismically upgraded at a moderate cost. It is also believed that the IAEA Benchmark study for seismic analysis and testing of WWER NPPs will develop recommendations to effective seismic upgrading of the existing plants

  20. Authorization Basis Safety Classification of Transfer Bay Bridge Crane at the 105-K Basins

    International Nuclear Information System (INIS)

    CHAFFEE, G.A.

    2000-01-01

    This supporting document provides the bases for the safety classification for the K Basin transfer bay bridge crane and the bases for the Structures, Systems, and Components (SSC) safety classification. A table is presented that delineates the safety significant components. This safety classification is based on a review of the Authorization Basis (AB). This Authorization Basis review was performed regarding AB and design baseline issues. The primary issues are: (1) What is the AB for the safety classification of the transfer bay bridge crane? (2) What does the SSC safety classification ''Safety Significant'' or ''Safety Significant for Design Only'' mean for design requirements and quality requirements for procurement, installation and maintenance (including replacement of parts) activities for the crane during its expected life time? The AB information on the crane was identified based on review of Department of Energy--Richland Office (RL) and Spent Nuclear Fuel (SNF) Project correspondence, K Basin Safety Analysis Report (SAR) and RL Safety Evaluation Reports (SERs) of SNF Project SAR submittals. The relevant correspondence, actions and activities taken and substantive directions or conclusions of these documents are provided in Appendix A

  1. Upgrading of the Budapest reactor

    International Nuclear Information System (INIS)

    Rosta, L.

    1986-10-01

    The increasing importance of neutron sources, high demand for irradiation and experimental facilities as well as improved safety requirements in the 'eighties, necessitate not only a technical modernization of the Reactor but an overall upgrade including instrumentation. Such a reconstruction was decided by the Hungarian governement in 1983

  2. DARHT: INTEGRATION OF AUTHORIZATION BASIS REQUIREMENTS AND WORKER SAFETY

    International Nuclear Information System (INIS)

    MC CLURE, D. A.; NELSON, C. A.; BOUDRIE, R. L.

    2001-01-01

    This document describes the results of consensus agreements reached by the DARHT Safety Planning Team during the development of the update of the DARHT Safety Analysis Document (SAD). The SAD is one of the Authorization Basis (AB) Documents required by the Department prior to granting approval to operate the DARHT Facility. The DARHT Safety Planning Team is lead by Mr. Joel A. Baca of the Department of Energy Albuquerque Operations Office (DOE/AL). Team membership is drawn from the Department of Energy Albuquerque Operations Office, the Department of Energy Los Alamos Area Office (DOE/LAAO), and several divisions of the Los Alamos National Laboratory. Revision 1 of the DARHT SAD had been written as part of the process for gaining approval to operate the Phase 1 (First Axis) Accelerator. Early in the planning stage for the required update of the SAD for the approval to operate both Phase 1 and Phase 2 (First Axis and Second Axis) DARHT Accelerator, it was discovered that a conflict existed between the Laboratory approach to describing the management of facility and worker safety

  3. Shielding for the upgraded duke free electron laser laboratory

    International Nuclear Information System (INIS)

    Vylet, V.

    2005-01-01

    The Duke FEL Laboratory is undergoing a series of upgrades staggered over time that will greatly increase the capabilities of the machines and by the same token the importance of radiation safety issues. In this paper, we present the scope of the planned upgrades and provide several specific examples of shielding calculations. We also present our effort to correlate calculations with experimental measurements. (authors)

  4. 10 CFR Appendix A to Subpart B of... - General Statement of Safety Basis Policy

    Science.gov (United States)

    2010-01-01

    ... at all levels. Performing work in accordance with the safety basis for a nuclear facility is the..., safety, and health into work planning and execution (48 CFR 970.5223-1, Integration of Environment, Safety and Health into Work Planning and Execution) and the DEAR clause on laws, regulations, and DOE...

  5. Core exit thermocouple upgrade at Zion station

    International Nuclear Information System (INIS)

    Ulinski, T.M.; Ferg, D.A.

    1989-01-01

    Following the Three Mile Island accident, the ability of the core exit thermocouple (CET) system to monitor reactor core conditions and core cooling status became a requirement of the U.S. Nuclear Regulatory Commission (NRC). Since the thermocouple system at Zion station was not originally required for postaccident monitoring, Commonwealth Edison Company (CECo) committed to upgrading the CET system and to installing a subcooling margin monitoring (SMM) system. The significance of this commitment was that CECo proposed to accomplish the upgrade effort using internal resources and by developing the required in-house expertise instead of procuring integrated packages from several nuclear steam supply system vendors. The result was that CECo was able to demonstrate a number of new capabilities and unique design features with a significant cost savings. These included a qualified connector with an integral thermocouple cold-reference junction temperature compensation; the design, assembly, testing, and installation of a seismically qualified class 1E microprocessor; a commercial-grade dedication/upgrade process for safety-related hardware; a human factors review capability, and a verification and validation program for safety-related software. A discussion of these new capabilities and details of the design features is presented in this paper

  6. Environmental upgrading of a landfill

    International Nuclear Information System (INIS)

    Agostinetto, V.; Vendrame, G.

    1999-01-01

    This article refers to an experimental study concerning the vegetative upgrading of a closed-down landfill (once used for industrial waste disposal). The aim was to check the possibility of reconstructing or aiding the natural growth of a vegetation in keeping with the surrounding area, in a tried environment such as that of landfills. The original idea contained in the approved project - which meant to generically upgrade the territory by planting species belonging to the grassy layer, shrubs and trees - has, with time, undergone some changes. On the basis of both the knowledge acquired during management and of a more accurate analysis of the territory, the experiment was preferred to aim at finding out which were the species, both continental and Mediterranean, able to gradually adjust to the surrounding landscape, leaving to natural selection the task to decide which species were more suitable to the upgrading of closed-down landfills, and which planting technique was more effective [it

  7. Report of a consultants' meeting on insights from PSA results on the programmes for safety upgrading of WWER NPPs. Extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    1996-11-01

    The objective of the meeting was to compare the insight from plant specific PSA studies with the safety upgrading programme of WWER NPPs. The PSAs were reviewed considering the scope, level and detail of PSA models and results of IAEA peer reviews. Safety improvements which are not normally included in PSAs were also considered. The review specifically considers for each plant specific PSA: the dominant initiating events and accident sequences contributing to core damage; and, the importance of systems, components and human actions to be used for prioritizing actions. 4 refs, tabs

  8. Seismic assessment and upgrading of Paks nuclear power plant

    International Nuclear Information System (INIS)

    Tamas, K.

    2001-01-01

    A comprehensive programme for seismic assessment and upgrading is currently in progress at Hungary's Paks NPP. The re-evaluation of the site seismic hazard had been already completed. The technology of safe shut down and heat removal is established and the systems and structures relevant for seismic safety are identified. A seismic instrumentation is installed. The pre-earthquake preparedness and post-earthquake actions are elaborated. The methods for seismic capacity assessment are selected. The seismic capacity evaluation and the design of upgrading measures are currently in progress. The easy to perform upgrading covering the most urgent measures had been already performed. (author)

  9. Modernization and power upgrading of the Loviisa NPP

    Energy Technology Data Exchange (ETDEWEB)

    Keskinen, A. [IVO Power Engineering Ltd., Vantaa (Finland)

    1997-12-31

    In 1995, Imatran Voima Oy (IVO) started a project for modernization and power upgrading of the Loviisa NPP. The main objectives of the project are to ensure plant safety, to increase electricity production and to improve the expertise of the IVO staff. The total electricity output of Loviisa 1 and 2 units is planned to be increased by about 100 MW. This will be achieved through renovation of the steam turbines and through gradual increase in the thermal reactor power up to 1,500 MW from the present level of 1,375 MW. The Loviisa NPP Final Safety Analysis Report has been revised to a great extent in connection with the licensing process of the reactor power upgrading. The project also includes certain improvements in the primary and safety systems to ensure plant safety. The total cost estimate of the project is around 200 million FIM. The project implementation started in 1995 and in accordance with the plans in 2000 after several phases the last measures at power plant will be completed. (orig.). 4 refs.

  10. Modernization and power upgrading of the Loviisa NPP

    International Nuclear Information System (INIS)

    Keskinen, A.

    1997-01-01

    In 1995, Imatran Voima Oy (IVO) started a project for modernization and power upgrading of the Loviisa NPP. The main objectives of the project are to ensure plant safety, to increase electricity production and to improve the expertise of the IVO staff. The total electricity output of Loviisa 1 and 2 units is planned to be increased by about 100 MW. This will be achieved through renovation of the steam turbines and through gradual increase in the thermal reactor power up to 1,500 MW from the present level of 1,375 MW. The Loviisa NPP Final Safety Analysis Report has been revised to a great extent in connection with the licensing process of the reactor power upgrading. The project also includes certain improvements in the primary and safety systems to ensure plant safety. The total cost estimate of the project is around 200 million FIM. The project implementation started in 1995 and in accordance with the plans in 2000 after several phases the last measures at power plant will be completed. (orig.)

  11. Modernization and power upgrading of the Loviisa NPP

    Energy Technology Data Exchange (ETDEWEB)

    Keskinen, A [IVO Power Engineering Ltd., Vantaa (Finland)

    1998-12-31

    In 1995, Imatran Voima Oy (IVO) started a project for modernization and power upgrading of the Loviisa NPP. The main objectives of the project are to ensure plant safety, to increase electricity production and to improve the expertise of the IVO staff. The total electricity output of Loviisa 1 and 2 units is planned to be increased by about 100 MW. This will be achieved through renovation of the steam turbines and through gradual increase in the thermal reactor power up to 1,500 MW from the present level of 1,375 MW. The Loviisa NPP Final Safety Analysis Report has been revised to a great extent in connection with the licensing process of the reactor power upgrading. The project also includes certain improvements in the primary and safety systems to ensure plant safety. The total cost estimate of the project is around 200 million FIM. The project implementation started in 1995 and in accordance with the plans in 2000 after several phases the last measures at power plant will be completed. (orig.). 4 refs.

  12. Seismic upgrading of the spent fuel storage building at Kozloduy NPP

    International Nuclear Information System (INIS)

    Alexandrov, A.; Borov, V.; Jordanov, M.; Karamanski, T.; Mihaylov, K.

    2001-01-01

    The Spent Fuel Storage Building at Kozloduy NPP site has been analysed for new review level earthquake with 0.2 g peak ground acceleration (compared to the initial design basis earthquake with 0.1 g PGA). The preliminary seismic analysis of the existing building structure using the 5% site specific response spectrum showed the need of seismic structural upgrading. Two upgrading concepts were evaluated on the basis of several factors. The main factor considered was preventing the collapse of the hall structure and the travelling cranes on the fuel storage area during and after a SSE. A three dimensional finite element model was created for the investigation of the seismic response of the existing structure and for the design of the building upgrading. The modelling of the heavy travelling crane and its sub-crane structure was one of the key points. Different configurations of the new upgrading and strengthening structures were investigated. Some interesting conclusions have been drawn from the experience in analysing and upgrading of such a complex industrial structure, comprised of elements with substantial differences in material, rigidity, construction and general behaviour. (author)

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

    International Nuclear Information System (INIS)

    Tomberlin, T.A.

    2002-01-01

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

  14. National Spherical Torus Experiment (NSTX) Center Stack Upgrade

    International Nuclear Information System (INIS)

    Neumeyer, C.; Avasarala, S.; Chrzanowski, J.; Dudek, L.; Fan, H.; Hatcher, H.; Heitzenroeder, P.; Menard, J.; Ono, M.; Ramakrishnan, S.; Titus, P.; Woolley, R.; Zhan, H.

    2009-01-01

    The purpose of the NSTX Center Stack Upgrade project is to expand the NSTX operational space and thereby the physics basis for next-step ST facilities. The plasma aspect ratio (ratio of plasma major to minor radius) of the upgrade is increased to 1.5 from the original value of 1.26, which increases the cross sectional area of the center stack by a factor of ∼ 3 and makes possible higher levels of performance and pulse duration.

  15. Computer codes used during upgrading activities at MINT TRIGA reactor

    International Nuclear Information System (INIS)

    Mohammad Suhaimi Kassim; Adnan Bokhari; Mohd Idris Taib

    1999-01-01

    MINT TRIGA Reactor is a 1-MW swimming pool nuclear research reactor commissioned in 1982. In 1993, a project was initiated to upgrade the thermal power to 2 MW. The IAEA assistance was sought to assist the various activities relevant to an upgrading exercise. For neutronics calculations, the IAEA has provided expert assistance to introduce the WIMS code, TRIGAP, and EXTERMINATOR2. For thermal-hydraulics calculations, PARET and RELAP5 were introduced. Shielding codes include ANISN and MERCURE. However, in the middle of 1997, MINT has decided to change the scope of the project to safety upgrading of the MINT Reactor. This paper describes some of the activities carried out during the upgrading process. (author)

  16. Working group 1A - basis for the standard-safety

    International Nuclear Information System (INIS)

    Whipple, C.

    1993-01-01

    This paper presents a summary of the progress made by working group 1A (Basis for the Safety Standard) during the Electric Power Research Institute's EPRI Workshop on the technical basis of EPA HLW Disposal Criteria, March 1993. This group discussed the semantics of terms within the standard 40 CFR Part 191, the implementation of this standard, the advanced notice of rulemaking, the issue of emitting carbon-14 through a gaseous pathway, the strategy of dealing with standards for contamination of drinking water and groundwater, the 100,000 year time frame, and the analysis of specific comments. The specific comments dealt with the cost effectiveness of the standard, the dose histogram for populations and individuals, groundwater definition and the underlying technology driver for this standard

  17. Snowflake divertor configuration studies for NSTX-Upgrade

    International Nuclear Information System (INIS)

    Soukhanovskii, V.A.

    2011-01-01

    Snowflake divertor experiments in NSTX provide basis for PMI development toward NSTX-Upgrade. Snowflake configuration formation was followed by radiative detachment. Significant reduction of steady-state divertor heat flux observed in snowflake divertor. Impulsive heat loads due to Type I ELMs are partially mitigated in snowflake divertor. Magnetic control of snowflake divertor configuration is being developed. Plasma material interface development is critical for NSTX-U success. Four divertor coils should enable flexibility in boundary shaping and control in NSTX-U. Snowflake divertor experiments in NSTX provide good basis for PMI development in NSTX-Upgrade. FY 2009-2010 snowflake divertor experiments in NSTX: (1) Helped understand control of magnetic properties; (2) Core H-mode confinement unchanged; (3) Core and edge carbon concentration reduced; and (4) Divertor heat flux significantly reduced - (a) Steady-state reduction due to geometry and radiative detachment, (b) Encouraging results for transient heat flux handling, (c) Combined with impurity-seeded radiative divertor. Outlook for snowflake divertor in NSTX-Upgrade: (1) 2D fluid modeling of snowflake divertor properties scaling - (a) Edge and divertor transport, radiation, detachment threshold, (b) Compatibility with cryo-pump and lithium conditioning; (2) Magnetic control development; and (3) PFC development - PFC alignment and PFC material choice.

  18. Thermal Upgrading of 9977 Radioactive Material (Ram) Type B Package

    International Nuclear Information System (INIS)

    Gupta, N.; Abramczyk, G.

    2012-01-01

    The 9977 package is a radioactive material package that was originally certified to ship Heat Sources and RTG contents up to 19 watts and it is now being reviewed to significantly expand its contents in support of additional DOE missions. Thermal upgrading will be accomplished by employing stacked 3013 containers, a 3013 aluminum spacer and an external aluminum sleeve for enhanced heat transfer. The 7th Addendum to the original 9977 package Safety Basis Report describing these modifications is under review for the DOE certification. The analyses described in this paper show that this well-designed and conservatively analyzed package can be upgraded to carry contents with decay heat up to 38 watts with some simple design modifications. The Model 9977 package has been designed as a replacement for the Department of Transportation (DOT) Fissile Specification 6M package. The 9977 package is a very versatile Type B package which is certified to transport and store a wide spectrum of radioactive materials. The package was analyzed quite conservatively to increase its usefulness and store different payload configurations. Its versatility is evident from several daughter packages such as the 9978 and H1700, and several addendums where the payloads have been modified to suit the Shipper's needs without additional testing.

  19. Discussion on the Criterion for the Safety Certification Basis Compilation - Brazilian Space Program Case

    Science.gov (United States)

    Niwa, M.; Alves, N. C.; Caetano, A. O.; Andrade, N. S. O.

    2012-01-01

    The recent advent of the commercial launch and re- entry activities, for promoting the expansion of human access to space for tourism and hypersonic travel, in the already complex ambience of the global space activities, brought additional difficulties over the development of a harmonized framework of international safety rules. In the present work, with the purpose of providing some complementary elements for global safety rule development, the certification-related activities conducted in the Brazilian space program are depicted and discussed, focusing mainly on the criterion for certification basis compilation. The results suggest that the composition of a certification basis with the preferential use of internationally-recognized standards, as is the case of ISO standards, can be a first step toward the development of an international safety regulation for commercial space activities.

  20. Recent Upgrade of the Klystron Modulator at SLAC

    International Nuclear Information System (INIS)

    Nguyen, M.N.; Burkhart, C.P.; Lam, B.K.; Morris, B.

    2011-01-01

    The SLAC National Accelerator Laboratory employs 244 klystron modulators on its two-mile-long linear accelerator that has been operational since the early days of the SLAC establishment in the sixties. Each of these original modulators was designed to provide 250 kV, 262 A and 3.5 μS at up to 360 pps using an inductance-capacitance resonant charging system, a modified type-E pulse-forming network (PFN), and a pulse transformer. The modulator internal control comprised of large step-start resistor-contactors, vacuum-tube amplifiers, and 120 Vac relays for logical signals. A major, power-component-only upgrade, which began in 1983 to accommodate the required beam energy of the SLAC Linear Collider (SLC) project, raised the modulator peak output capacity to 360 kV, 420 A and 5.0 μS at a reduced pulse repetition rate of 120 pps. In an effort to improve safety, performance, reliability and maintainability of the modulator, this recent upgrade focuses on the remaining three-phase AC power input and modulator controls. The upgrade includes the utilization of primary SCR phase control rectifiers, integrated fault protection and voltage regulation circuitries, and programmable logic controllers (PLC) -- with an emphasis on component physical layouts for safety and maintainability concerns. In this paper, we will describe the design and implementation of each upgraded component in the modulator control system. We will also report the testing and present status of the modified modulators.

  1. System Safety Program Plan for Project W-314, tank farm restoration and safe operations

    International Nuclear Information System (INIS)

    Boos, K.A.

    1996-01-01

    This System Safety Program Plan (SSPP) outlines the safety analysis strategy for project W-314, ''Tank Farm Restoration and Safe Operations.'' Project W-314 will provide capital improvements to Hanford's existing Tank Farm facilities, with particular emphasis on infrastructure systems supporting safe operation of the double-shell activities related to the project's conceptual Design Phase, but is planned to be updated and maintained as a ''living document'' throughout the life of the project to reflect the current safety analysis planning for the Tank Farm Restoration and Safe Operations upgrades. This approved W-314 SSPP provides the basis for preparation/approval of all safety analysis documentation needed to support the project

  2. Upgrading of seismic design of nuclear power plant building

    Energy Technology Data Exchange (ETDEWEB)

    Akiyama, Hiroshi [Tokyo Univ. (Japan). Faculty of Engineering; Kitada, Yoshio

    1997-03-01

    In Japan seismic design methodology of nuclear power plant (NPP) structures has been established as introduced in the previous session. And yet efforts have been continued to date to upgrade the methodology, because of conservative nature given to the methodology in regard to unknown phenomena and technically-limited modeling involved in design analyses. The conservative nature tends to produce excessive safety margins, and inevitably send NPP construction cost up. Moreover, excessive seismic design can increase the burden on normal plant operation, though not necessarily contributing to overall plant safety. Therefore, seismic engineering has put to many tests and simulation analyses in hopes to rationalize seismic design and enhance reliability of seismic safety of NPPs. In this paper, we describe some studies on structural seismic design of NPP underway as part of Japan`s effort to upgrade existing seismic design methodology. Most studies described here are carried out by NUPEC (Nuclear Power Engineering Company) funded by MITI (the Ministry of International Trade and Industry Japan), though, similar studies with the same motive are also carrying out by nuclear industries such as utilities, NPP equipment and system manufacturers and building constructors. This paper consists of three sections, each introducing studies relating to NPP structural seismic design, new siting technology, and upgrading of the methodology of structural design analyses. (J.P.N.)

  3. Upgrading of seismic design of nuclear power plant building

    International Nuclear Information System (INIS)

    Akiyama, Hiroshi; Kitada, Yoshio.

    1997-01-01

    In Japan seismic design methodology of nuclear power plant (NPP) structures has been established as introduced in the previous session. And yet efforts have been continued to date to upgrade the methodology, because of conservative nature given to the methodology in regard to unknown phenomena and technically-limited modeling involved in design analyses. The conservative nature tends to produce excessive safety margins, and inevitably send NPP construction cost up. Moreover, excessive seismic design can increase the burden on normal plant operation, though not necessarily contributing to overall plant safety. Therefore, seismic engineering has put to many tests and simulation analyses in hopes to rationalize seismic design and enhance reliability of seismic safety of NPPs. In this paper, we describe some studies on structural seismic design of NPP underway as part of Japan's effort to upgrade existing seismic design methodology. Most studies described here are carried out by NUPEC (Nuclear Power Engineering Company) funded by MITI (the Ministry of International Trade and Industry Japan), though, similar studies with the same motive are also carrying out by nuclear industries such as utilities, NPP equipment and system manufacturers and building constructors. This paper consists of three sections, each introducing studies relating to NPP structural seismic design, new siting technology, and upgrading of the methodology of structural design analyses. (J.P.N.)

  4. NPP Krsko Severe Accident Management Guidelines Upgrade

    International Nuclear Information System (INIS)

    Mihalina, Mario; Spalj, Srdjan; Glaser, Bruno; Jalovec, Robi; Jankovic, Gordan

    2014-01-01

    Nuclear Power Plant Krsko (NEK) has decided to take steps for upgrade of safety measures to prevent severe accidents, and to improve the means to successfully mitigate their consequences. The content of the program for the NEK Safety Upgrade is consistent with the nuclear industry response to Fukushima accident, which revealed many new insights into severe accidents. Therefore, new strategies and usage of new systems and components should be integrated into current NEK Severe Accident Management Guidelines (SAMG's). SAMG's are developed to arrest the progression of a core damage accident and to limit the extent of resulting releases of fission products. NEK new SAMG's revision major changes are made due to: replacement of Electrical Recombiners by Passive Autocatalytic Recombiners (PARs) and the installation of Passive Containment Filtered Vent System (PCFV); to handle a fuel damage situation in Spent Fuel Pool (SFP) and to assess risk of core damage situation during shutdown operation. (authors)

  5. Pakistan upgrades PARR-1 and converts to LEU

    International Nuclear Information System (INIS)

    Anon.

    1990-01-01

    The Pakistan Research Reactor, PARR-1, is a 5MW swimming pool type reactor originally designed to use MTR type fuel elements fabricated from uranium enriched to more than 90%. After about 24 years of satisfactory operation it is now planned to convert the reactor to use low enriched (20%) uranium fuel. The opportunity will also be taken to upgrade the reactor power to about 9MW. This power upgrading will meet the demand for higher neutron fluxes for experimental and radioisotope production as well as compensating for the neutron flux penalty arising from conversion from high enriched to low enriched fuel. During the process of conversion and upgrading it is also proposed to renovate existing services and associated systems and to add certain new safety related engineering. (author)

  6. WWER-440/V-230 Confinement modernization to upgrade the critical safety function 'Containment integrity' in case of severe accident

    International Nuclear Information System (INIS)

    Sartmadjiev, A.

    1999-01-01

    In this lecture the WWER-440/V-230 confinement modernization to upgrade the critical safety function 'Containment integrity' in case of severe accident is presented. There are discussed: design limitations of the location system; consequence from these design limitations; a few confinement reconstruction concepts of this type of units worldwide; and purpose of the confinement reconstruction - to improve significantly the original design, ensuring (1) localization for all possible primary breaks and (2) limitation of the radiological consequences for the personnel, the population and the environment below the regulatory requirements

  7. Risk Assessment on the Transition Program for Air Traffic Control Automation System Upgrade

    Directory of Open Access Journals (Sweden)

    Li Dong Bin

    2016-01-01

    Full Text Available We analyzed the safety risks of the transition program for Air Traffic Control (ATC automation system upgrade by using the event tree analysis method in this paper. We decomposed the occurrence progress of the three transition phase and built the event trees corresponding to the three stages, and then we determined the probability of success of each factor and calculated probability of success of the air traffic control automation system upgrade transition. In the conclusion, we illustrate the transition program safety risk according to the results.

  8. Materials Safety Data Sheets: the basis for control of toxic chemicals

    Energy Technology Data Exchange (ETDEWEB)

    Ketchen, E.E.; Porter, W.E.

    1979-09-01

    The Material Safety Data Sheets contained in this volume are the basis for the Toxic Chemical Control Program developed by the Industrial Hygiene Department, Health Division, ORNL. The three volumes are the update and expansion of ORNL/TM-5721 and ORNL/TM-5722 Material Safety Data Sheets: The Basis for Control of Toxic Chemicals, Volume I and Volume II. As such, they are a valuable adjunct to the data cards issued with specific chemicals. The chemicals are identified by name, stores catalog number where appropriate, and sequence numbers from the NIOSH Registry of Toxic Effects of Chemical Substances, 1977 Edition, if available. The data sheets were developed and compiled to aid in apprising the employees of hazards peculiar to the handling and/or use of specific toxic chemicals. Space limitation necessitate the use of descriptive medical terms and toxicological abbreviations. A glossary and an abbreviation list were developed to define some of those sometimes unfamiliar terms and abbreviations. The page numbers are keyed to the catalog number in the chemical stores at ORNL.

  9. UPGRADES

    CERN Multimedia

    J. Butler and J. Nash

    2011-01-01

    Recent progress on the CMS upgrades was summarised, in a workshop held at Fermilab between 7th and 10th November, attended by more than 150 people, many of whom came from Europe and Asia. Important goals of the workshop were to begin to formulate a schedule for the upgrades and to determine project interdependencies. Input was received from all the upgrade working groups and will be combined into a first-pass schedule over the next several weeks. In addition, technical progress on each of the major subtasks was presented and plans for the near-term future were established. Slides from the more than 100 talks are located at: https://indico.cern.ch/conferenceDisplay.py?confId=153564 In the opening plenary session, Frank Zimmermann, of the CERN Beams Department, gave his view of the LHC luminosity evolution. The luminosity will increase faster than we assumed in designing the upgrades. CMS will need to re-evaluate the current upgrade plans and revise them if necessary. CMS Upgrade Physics coordinator...

  10. Upgrade of VR-1 training reactor I and C

    International Nuclear Information System (INIS)

    Kropik, M.; Matejka, K.; Chab, V.

    2003-01-01

    The contribution describes the upgrade of the VR-1 training reactor I and C (Instrumentation and Control). The reactor was put into operation in the 1990, and its I and C seems to be obsolete now. The new I and C utilises the same digital technology as the old one. The upgrade has been done gradually during holidays in order not to disturb the reactor utilisation during teaching and training. The first stage consisted in the human-machine interface and the control room upgrade in 2001. A new operator's desk, displays, indicators and buttons were installed. Completely new software and communication interface to the present I and C were developed. During the second stage in 2002, new control rod drivers and safety circuits were installed. The rod motors were replaced and necessary mechanical changes on the control rod mechanism, induced by the utilisation of the new motor, were done. The new safety circuits utilise high quality relays with forced contacts to guarantee high reliability of their operation. The third stage, the control system upgrade is being carried out now. The new control system is based on an industrial PC mounted in a 19 inch crate. The operating system of the PC is the Microsoft Windows XP with the real time support RTX of the VentureCom Company. A large amount of work has been devoted to the software requirements to specify all dependencies, modes and permitted actions, safety measures, etc. The Department took an active part in the setting of software requirements and later in verification and validation of the software and the whole control system. Finally, a new protection system consisting of power measuring and power protection channels will be installed in 2004 or 2005. (author)

  11. I and C upgrading at nuclear power plants

    International Nuclear Information System (INIS)

    Tamiri, A.

    2003-01-01

    Continuing the operation of existing nuclear power plants will help reduce the number of new base-load nuclear and fossil power plants that need to be built. Old nuclear power plants in Canada are operating with analog instrumentation and control systems. For a number of reasons, such as changes and improvements in the applicable standards and design, maintenance problems due to the lack of spares, technical obsolescence, the need to increase power production, availability, reliability and safety, and in order to reduce operation and maintenance costs, instrumentation and control upgrading at nuclear power plants in a cost effective manner should be considered the greatest priority. Failures of instrumentation and control (I and C) due to aging and obsolescence issues may have an immediate negative impact on plant reliability and availability and also affect long-term plant performance and safety. In today's competitive marketplace, power plants are under pressure to cut spending on maintenance while reducing the risk of equipment failure that could cause unplanned outage. To improve plant safety and availability, old nuclear power plants will require investment in new technologies that can improve the performance and reduce the costs of generation by addressing the long term reliability of systems by up-grading to modem digital instrumentation and control and optimization opportunities. Boiler drum level control at nuclear power plants is critical for both plant protection and equipment safety and applies equality to high and low levels of water within the boiler drum. Plant outage studies at Pickering Nuclear have identified boiler drum level control and feed water control systems as major contributors to plant unavailability. Ways to improve transient and steady state response, upgrading existing poor analog control systems for boiler level and feed-water control systems at Pickering Nuclear, with enhanced and robust controller will be discussed in this paper

  12. Dukovany nuclear power plant safety

    International Nuclear Information System (INIS)

    1999-01-01

    Presentation covers recommended safety issues for the Dukovany NPP which have been solved with satisfactory conclusions. Safety issues concerned include: radiation safety; nuclear safety; security; emergency preparedness; health protection at work; fire protection; environmental protection; chemical safety; technical safety. Quality assurance programs at all stages on NPP life time is described. Report includes description of NPP staff training provision, training simulator, emergency operating procedures, emergency preparedness, Year 2000 problem, inspections and life time management. Description of Dukovany Plant Safety Analysis Projects including integrity of the equipment, modernisation, equipment innovation and safety upgrading program show that this approach corresponds to the actual practice applied in EU countries, and fulfilment of current IAEA requirements for safety enhancement of the WWER 440/213 units in the course of MORAWA Equipment Upgrading program

  13. Configuration management after design basis reconstitution

    International Nuclear Information System (INIS)

    Purcell, J.J.; Livingston, B.R.

    1991-01-01

    Over the last few years, Fort Calhoun station (FCS) has implemented a number of programs to enhance plant operability and readiness. The design basis document (DBD) reconstitution project was the cornerstone of this effort. Vendor manual upgrade, operating procedures upgrade, plant equipment data-base verification, equipment labeling, and warehousing improvements were also implemented as part of this improvement program. With the completion of these programs, plant documentation was current to the baselines established by each program, and a configuration management program (CMP) was established to maintain this level of accuracy throughout the remaining life of FCS. Change control throughout the organization has been reviewed and upgraded to ensure that all changes are evaluated for impact to the design bases

  14. Technical basis for instrumentation and control design improvements in WWER-440/230 nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    1996-01-01

    Instrumentation and control (I and C) has been recognized as an area which requires substantial improvements in WWER NNPs, particularly for model 230 plants. Under contract with the IAEA the Spanish company Empresarios Agrupados (EA) developed a basic document proposing a technical basis for improvements related to the following most significant aspects of I and C: criteria for safety classification; remote shutdown panel; I and C support to operation and control room design; instrumentation set point margins; accident monitoring instrumentation. This publication is derived from the original report of EA which was circulated by the IAEA for review by staff members and experts from various Member States. It was finally agreed upon at a Consultants' meeting convened by the IAEA in Vienna in May 1994 with the participation of experts from France, Germany and Spain. The guidance expressed in this report is based on the IAEA/NUSS standards, safety guides and practices, and on regulations in use in various Member States. It is proposed as a way of carrying out the necessary studies to improve safety by upgrading the vital part of instrumentation an control in WWER-440 model 230 nuclear power plants. 28 refs, 3 figs

  15. Fault tree analysis of Project S-4404, Upgrade Canyon Exhaust System

    International Nuclear Information System (INIS)

    Browne, E.V.; Low, J.M.; Lux, C.R.

    1992-01-01

    Project S-4404, Upgrade Canyon Exhaust Systems, is a $177 million project with the purpose of upgrading the Exhaust Systems for both F and H Canyon Facilities. This upgrade will replace major portions of the F and H-Canyon exhaust systems, downstream of their respective sand filters with higher capacity and more reliable systems. Because of the high cost, DOE requested Program Control ampersand Integration (PC ampersand I) to examine specific deletions to the project. PC ampersand I requested Nuclear Processes Safety Research (NPSR) to perform an analysis to compare failure rates for the existing F ampersand H Canyon exhaust systems with the proposed exhaust system and specific proposed exhaust system alternatives. The objective of this work was to perform an analysis and compare failure rates for the existing F ampersand H Canyon exhaust systems with the proposed project exhaust system and proposed project alternatives. Based on fault tree analysis, two conclusions are made. First, D ampersand D activities can be eliminated from the project with no significant decrease to exhaust system safety. Deletion of D ampersand D activities would result in a cost savings of $29 million. Second, deletion of DOE Order 6430.1A requirements regarding DBAs would decrease exhaust system safety by a factor of 12

  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. Experiences in seismic upgrading of equipment and structures in Kozloduy nuclear power plant (440 WWER-PWR)

    International Nuclear Information System (INIS)

    Ordonez Villalobos, A.

    1993-01-01

    Within the framework of the 'Emergency programme for Nuclear Safety of Kozloduy NPP' it has been concluded that the increase in seismic safety of a NPP can be achieved by upgrading the key equipment in a cost effective way. Essential and vulnerable equipment has to be identified. Seismic capacity should be evaluated base don realistic state of the art criteria. Seismic review teams ef experienced engineers should conduct planned walk-downs in order to propose effective upgrading solutions. Team work of plan engineers and construction engineers would enhance the effectiveness of the solutions. It is recommended that all the participants be motivated and have a clear understanding of the objectives of the upgrading

  18. Feasibility study for Tehran Research Reactor power upgrading

    Energy Technology Data Exchange (ETDEWEB)

    Farhadi, Kazem [Nuclear Research Center, Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of)], E-mail: kfarhadi@aeoi.org.ir; Khakshournia, Samad [Nuclear Research Center, Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of)

    2008-07-15

    The present work is concerned with a power upgrading study of Tehran Research Reactor (TRR). The upgrading study is aimed at investigating the possibility of raising power of the TRR from the current level of 5 MW{sub th} to a higher level without violating the original thermal-hydraulic safety criteria. The existing core, comprising 22 standard fuel elements and five control fuel elements, is used for the analyses. Different reactor thermal powers (5-11 MW) and different core coolant flow rates (500-921 m{sup 3}/h) are considered. It is shown that, for the present core, this goal could be achieved safely by gradually opening the butterfly control valve until the desired coolant flow rate is reached. The TRR power could be upgraded up to around 7.5 MW{sub th} with the total power peaking factor maintained at less than or equal to 3.0.

  19. Feasibility study for Tehran Research Reactor power upgrading

    International Nuclear Information System (INIS)

    Farhadi, Kazem; Khakshournia, Samad

    2008-01-01

    The present work is concerned with a power upgrading study of Tehran Research Reactor (TRR). The upgrading study is aimed at investigating the possibility of raising power of the TRR from the current level of 5 MW th to a higher level without violating the original thermal-hydraulic safety criteria. The existing core, comprising 22 standard fuel elements and five control fuel elements, is used for the analyses. Different reactor thermal powers (5-11 MW) and different core coolant flow rates (500-921 m 3 /h) are considered. It is shown that, for the present core, this goal could be achieved safely by gradually opening the butterfly control valve until the desired coolant flow rate is reached. The TRR power could be upgraded up to around 7.5 MW th with the total power peaking factor maintained at less than or equal to 3.0

  20. Safety improvement of Paks nuclear power plant

    International Nuclear Information System (INIS)

    Vamos, G.

    1999-01-01

    Safety upgrading completed in the early nineties at the Paks NPP include: replacement of steam generator safety valves and control valves; reliability improvement of the electrical supply system; modification of protection logic; enhancement of the fire protection; construction of full scope Training Simulator. Design safety upgrading measures achieved in recent years were concerned with: relocation of steam generator emergency feed-water supply; emergency gas removal from the primary coolant system; hydrogen management in the containment; protection against sumps; preventing of emergency core cooling system tanks from refilling. Increasing seismic resistance, containment assessment, refurbishment of reactor protection system, improving reliability of emergency electrical supply, analysis of internal hazards are now being implemented. Safety upgrading measures which are being prepared include: bleed and feed procedures; reactor over-pressurisation protection in cold state; treatment of steam generator primary to secondary leak accidents. Operational safety improvements are dealing with safety culture, training measures and facilities; symptom based emergency operating procedures; in-service inspection; fire protection. The significance of international cooperation is emphasised in view of achieving nuclear safety standards recognised in EU

  1. 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)

  2. 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)

  3. Scientific basis for a safety case of deep geological repositories

    Energy Technology Data Exchange (ETDEWEB)

    Noseck, Ulrich; Becker, Dirk-Alexander; Brasser, Thomas [and others

    2012-11-15

    Within this project strategies and methods to build a safety case for deep geological repositories are further developed. This includes also the scientific fundamentals as a basis of the safety case. In the international framework the methodology of the Safety Case is frequently applied and continuously improved. According to definitions from IAEA and NEA the Safety Case is a compilation of arguments and facts, which describe, quantify and support the safety and the degree of confidence in the safety of the geological repository. The safety of the geological repository should be demonstrated by the safety case. The safety case is the basis for essential decisions during a repository programme. It comprises the results of safety assessments in combination with additional information like multiple lines of evidence and a discussion of robustness and quality of the repository, its design and the quality of all safety assessments including the basic assumptions. A crucial element of the Safety Case is the long-term safety analysis, i.e. the systematic analysis of the hazards connected with the facility and the capability of site and repository design to ensure the required safety functions and to fulfill the technical claims. Long-term safety analysis requires a powerful and qualified programme package, which contains appropriate hardware and software as well as well trained and experienced modellers performing the model calculations. The calculation tools used within safety cases need to be checked and verified and continuously adapted to the state-of-the-art science and technology. Especially it needs to be applicable to a real repository system. For the assessment of safety a deep process understanding is necessary. The R and D work performed within this project will contribute to the improvement of process and system understanding as well as to the further development of methods and strategies applied in the safety case. Emphasis was put on the following aspects

  4. Scientific basis for a safety case of deep geological repositories

    International Nuclear Information System (INIS)

    Noseck, Ulrich; Becker, Dirk-Alexander; Brasser, Thomas

    2012-11-01

    Within this project strategies and methods to build a safety case for deep geological repositories are further developed. This includes also the scientific fundamentals as a basis of the safety case. In the international framework the methodology of the Safety Case is frequently applied and continuously improved. According to definitions from IAEA and NEA the Safety Case is a compilation of arguments and facts, which describe, quantify and support the safety and the degree of confidence in the safety of the geological repository. The safety of the geological repository should be demonstrated by the safety case. The safety case is the basis for essential decisions during a repository programme. It comprises the results of safety assessments in combination with additional information like multiple lines of evidence and a discussion of robustness and quality of the repository, its design and the quality of all safety assessments including the basic assumptions. A crucial element of the Safety Case is the long-term safety analysis, i.e. the systematic analysis of the hazards connected with the facility and the capability of site and repository design to ensure the required safety functions and to fulfill the technical claims. Long-term safety analysis requires a powerful and qualified programme package, which contains appropriate hardware and software as well as well trained and experienced modellers performing the model calculations. The calculation tools used within safety cases need to be checked and verified and continuously adapted to the state-of-the-art science and technology. Especially it needs to be applicable to a real repository system. For the assessment of safety a deep process understanding is necessary. The R and D work performed within this project will contribute to the improvement of process and system understanding as well as to the further development of methods and strategies applied in the safety case. Emphasis was put on the following aspects

  5. Siemens tasks and activities during gradual upgrading

    International Nuclear Information System (INIS)

    Kuehne, B.

    2001-01-01

    The gradual upgrading of NPP V1 Bohunice constitutes the most extensive reconstruction project to be performed in an operational WWER plant to date. It constitutes a milestone in Siemens co-operation on projects for WWERs to date and is an encouraging example for the successful combination of western and eastern safety cultures. During the runtime of this project Siemens learnt very much about the 'complete function' as well as the details of this WWER plant type. In contrast the Slovakian partners really gained a much better understanding for the western approach to ensure safety, for the most modern safety technology, in particular for instrumentation and control, and last but not least or advanced project management. (author)

  6. Upgrading of the research reactors FRG-1 and FRG-2

    International Nuclear Information System (INIS)

    Krull, W.

    1981-01-01

    In 1972 for the research reactor FRG-2 we applied for a license to increase the power from 15 MW to 21 MW. During this procedure a public laying out of the safety report and an upgrading procedure for both research reactors - FRG-1 (5 MW) and FRG-2 - were required by the licensing authorities. After discussing the legal background for licensing procedures in the Federal Republic of Germany the upgrading for both research reactors is described. The present status and future licensing aspects for changes of our research reactors are discussed, too. (orig.) [de

  7. Upgrading Planning and Executive Strategy for Reactor Protection System and Relative Equipment in Qinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Jiang Zuyue

    2010-01-01

    Qinshan Nuclear Power Plant (QNPP) is the first nuclear power plant in China which completed the reactor protection system (RPS) upgrading with new digital safety instrumentation and control (I and C) platform instead of original analog system. At the same time,the nuclear instrumentation system (NIS) was upgraded with the same digital I and C platform. For adapting QNPP's actual engineering situation,the upgrading planning was taken by comprehensively investigating current development and application of digital safety I and C platform in the worldwide scope and by reviewing plant's original systems operation history. The project executive strategy-QNPP's leading role with necessary overseas cooperation and internal technical supports as great as possible, was determined. Some significant factors might influence and restrict the RPS and relative equipment upgrading executive actions in an operating NPP were analyzed.Finally, the engineering feasibility was briefly assessed to recognize the anticipated issues and difficulties and to prepare the relative solutions in advance for the purpose of ensuring the RPS upgrading objectives completely realized. (authors)

  8. 14 CFR 60.37 - FSTD qualification on the basis of a Bilateral Aviation Safety Agreement (BASA).

    Science.gov (United States)

    2010-01-01

    ... Bilateral Aviation Safety Agreement (BASA). 60.37 Section 60.37 Aeronautics and Space FEDERAL AVIATION... CONTINUING QUALIFICATION AND USE § 60.37 FSTD qualification on the basis of a Bilateral Aviation Safety... on International Civil Aviation for the sponsor of an FSTD located in that contracting State may be...

  9. Research program of the high temperature engineering test reactor for upgrading the HTGR technology

    International Nuclear Information System (INIS)

    Kunitomi, Kazuhiko; Tachibana, Yukio; Takeda, Takeshi; Saikusa, Akio; Sawa, Kazuhiro

    1997-07-01

    The High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated and helium-cooled reactor with an outlet power of 30 MW and outlet coolant temperature of 950degC, and its first criticality will be attained at the end of 1997. In the HTTR, researches establishing and upgrading the technology basis necessary for an HTGR and innovative basic researches for a high temperature engineering will be conducted. A research program of the HTTR for upgrading the technology basis for the HTGR was determined considering realization of future generation commercial HTGRs. This paper describes a research program of the HTTR. (author)

  10. Seismic safety review mission to assist in the evaluation of the design of seismic upgrading for Kozloduy NPP. Sofia, Bulgaria, 19-23 October 1992

    International Nuclear Information System (INIS)

    Ma, D.; Prato, C.; Godoy, A.

    1992-10-01

    A seismic Safety Review Mission to assist in the evaluation of the design of seismic upgrading for Kozloduy NPP was performed in Sofia from 19-23 October 1992. The objectives of the mission were to assist the Bulgarian authorities in: the evaluation of the floor response spectra of the main buildings of units 1-4 at Kozloduy NPP, calculated for the new defined seismic parameters at site (Review Level Earthquake - RLE); the evaluation of the remedial and strengthening measures proposed for the seismic upgrading of the pump house and diesel generator buildings to the new defined RLE. This mission completed the scope of previous IAEA mission - BUL/9/012-18b - (see Report 3262) performed from 3-7 August 1992, with regard to tasks which were not evaluated at that time because they had not been finished. 2 tabs

  11. Nuclear safety in Slovak Republic. Status of safety improvements

    International Nuclear Information System (INIS)

    Toth, A.

    1999-01-01

    Status of the safety improvements at Bohunice V-1 units concerning WWER-440/V-230 design upgrading were as follows: supplementing of steam generator super-emergency feed water system; higher capacity of emergency core cooling system; supplementing of automatic links between primary and secondary circuit systems; higher level of secondary system automation. The goal of the modernization program for Bohunice V-1 units WWER-440/V-230 was to increase nuclear safety to the level of the proposals and IAEA recommendations and to reach probability goals of the reactor concerning active zone damage, leak of radioactive materials, failures of safety systems and damage shields. Upgrading program for Mochovce NPP - WWER-440/V-213 is concerned with improving the integrity of the reactor pressure vessel, steam generators 'leak before break' methods applied for the NPP, instrumentation and control of safety systems, diagnostic systems, replacement of in-core monitoring system, emergency analyses, pressurizers safety relief valves, hydrogen removal system, seismic evaluations, non-destructive testing, fire protection. Implementation of quality assurance has a special role in improvement of operational safety activities as well as safety management and safety culture, radiation protection, decommissioning and waste management and training. The Year 2000 problem is mentioned as well

  12. Role of computer techniques for knowledge propagation about nuclear energetics safety

    International Nuclear Information System (INIS)

    Osachkin, V.S.

    1996-01-01

    The development of nuclear power engineering depends on the levels of nuclear, radiological and ecological safety. To ensure the approval of such levels by the community to spread the knowledge on Safety of Nuclear Engineering in understandable forms. New computer technologies may play an important role in the safety education of the public and upgrading of qualification of personnel. The progress in computer nets development makes it possible to use besides e-mail qualification of personnel. The progress in computer in nets development makes it possible to use besides e-mail and BBS the Internet system for remote education. As an example a computer course on Atomic Energy and its safety presented. This course now written in Russian consists of 6 parts, namely: physical basis of utilization of Nuclear energy; technical bases of uses of Nuclear energy; nuclear Reactors and their Systems; safety Principles, Goals, Nuclear Safety Regulation; the Environmental Impact of the us of Nuclear Power, severe accident consequences and scenarios

  13. Upgrading and modernization of the high flux reactor Petten

    International Nuclear Information System (INIS)

    Ahlf, J.

    1992-01-01

    The High Flux Reactor (HFR) at Petten, The Netherlands, owned by the European Communities and operated by the Netherlands Energy Research Foundation, is a water-cooled and moderated, multipurpose research reactor of the closed-tank in-pool type, operated at 45 MW. Performance upgrading comprised two power increases from 20 MW via 30 MW to 45 MW, providing more and higher rated irradiation positions in the tank. With the replacement of the original reactor vessel the experimental capabilities of the reactor were improved. Better pool side facilities and the introduction of a large cross-section, double, beam tube were implemented. Additional major installation upgrading activities consisted of the replacement of the primary and the pool heat exchangers, replacement of the beryllium reflector elements, extension of the overpower protection systems and upgrading of the nuclear instrumentation as well as the guaranteed power supply. Control room upgrading is in progress. A full new safety analysis, as well as the introduction of a comprehensive Quality Assurance system, are summarized under software upgrading. Continuous modernization and upgrading also takes place of equipment for fuel and structural materials irradiations for fission reactors and future fusion machines. In parallel, all supporting services, as well as the management structure for large irradiation programmes, have been developed. Presently the reactor is operating at about 275 full power days per year with an average utilization of the irradiation positions of 70 to 80%. (orig.)

  14. B physics with the CDF Run II upgrade

    International Nuclear Information System (INIS)

    DeJongh, F.

    1996-01-01

    During the Run 1 data taking period, from 1992 through 1995, CDF has acquired 110 pb -1 of p anti p collisions at a center of mass energy of 1,800 GeV. This data has provided many results on B physics, and provides a basis for extrapolating to Run 2, which is scheduled to start in 1999 after major upgrades to both the accelerator and detector. The authors present herein a summary of Run 1 results relevant to an analysis of the CP asymmetry in B → J/ψK s , the CDF upgrade plans for Run 2, and some of the main B physics goals related to the exploration of the origin of CP violation

  15. General safety basis development guidance for environmental restoration decontamination and decommissioning

    International Nuclear Information System (INIS)

    Ellingson, D.R.; Kerr, N.; Bohlander, K.; Hansen, J.; Crowley, W.

    1994-02-01

    Safety analyses have the objective of contributing to two essential ingredients of a successful operation. The first is promoting the safety of the operation through worker involvement in information development (safety basis). The second is obtaining approval to conduct the operation (authorization). Typically these ingredients are assembled under separate programs covered by separate DOE requirements. DOE authorization relies on successful development of a document containing up to 21 topics written in terms and language suited to reviewers and approvers. Safety relies on successful training and procedures that convert the technical documented information into terms and language understandable to the worker. This separation can lead to successful incorporation of one ingredient independent of the other. At best, this separation may result in a safe but unauthorized operation; at worst, the separation may result in an unsafe operation authorized to proceed. This guide is based on experiences gained by contractors who have integrated rather than separated the safety and authorization. The short duration of ER/D ampersand D activities, the uncertainties of hazards, and the publicly expressed desire for demonstrable progress in cleanup activities add emphasis to the need to integrate rather than separate and develop new programs. Experience-based information has been useful to workers, safety analysis practitioners, and reviewers in the following ways: (1) Acquiring or developing the needed information in a useful form; (2) Managing the uncertainties during activity development and operation; (3) Identifying the subset of applicable requirements for an activity; (4) Developing the appropriate level of documentation detail for a specific activity; and (5) Increasing the usefulness and use of safety analysis (ownership)

  16. Upgrade of the CMS hardron calorimeter for an upgraded LHC

    OpenAIRE

    Anderson, Jake

    2012-01-01

    The CMS barrel and endcap hadron calorimeters (Hcal) upgrading the current photo-sensors are hybrid photodiodes (HPDs) to meet the demands of the upgraded luminosity of the LHC. A key aspect of the Hcal upgrade is to add longitudinal segmentation to improve background rejection, energy resolution, and electron isolation at L1 trigger. The increased segmentation can be achieved by replacing the HPD's with multi-pixel Geiger-mode avalanche photodiodes. The upgraded electron...

  17. Progress in MPC and A upgrades at Luch

    International Nuclear Information System (INIS)

    Mizin, P.; Chukov, V.; Rogatchev, V.; Curtiss, J.; Erkkila, B.; Goodey, K.; Hembree, D. Jr.; Lowe, D.; Turner, C.

    1997-01-01

    Luch, a MINATOM facility, has been engaged in both scientific research and uranium processing for fifty years. Since the spring of 1996, Luch has participated in a program of US/Russia Cooperation in Nuclear MPC and A Upgrades. The program began with planning for immediate upgrades in MPC and A, with en emphasis on physical protection. In addition, US and Luch experts exchanged technical data during a number of workshops, to establish a common understanding of available MPC and A tools and equipment. Site characterizations and vulnerability assessments were then prepared by Luch, to form the basis for the current program of methodical upgrades in all areas of MPC and A. Access control, alarms and alarm communications are being improved as part of this program. Control of nuclear material is being enhanced through improvements in material monitoring and in transportation security when nuclear material is moved between buildings on the Luch site. A comprehensive, site-wide computer network for Luch was designed during a recent workshop. Acquiring and installing this computer system, complete with COREMAS software, is currently in progress. Nuclear material analysis will be improved through NDA techniques using Canberra InSpector systems. The planned upgrades in nuclear MPC and A will reinforce safeguards over large quantities of HEU at Luch

  18. HOM Survey of the First CEBAF Upgrade Style Cavity Pair

    International Nuclear Information System (INIS)

    Marhauser, Frank; Daly, Edward; Davis, G.; Drury, Michael; Grenoble, Christiana; Hogan, John; Manus, Robert; Preble, Joseph; Reece, Charles; Rimmer, Robert; Tian, Kai; Wang, Haipeng

    2009-01-01

    The planned upgrade of the Continuous Electron Beam Accelerator Facility (CEBAF) at the Thomas Jefferson National Accelerator Laboratory (JLab) requires ten new superconducting rf (SRF) cavity cryomodules to double the beam energy to the envisaged 12 GeV. Adequate cavity Higher Order Mode (HOM) suppression is essential to avoid multipass, multibunch beam break-up (BBU) instabilities of the recirculating beam. We report on detailed HOM surveys performed for the first two upgrade style cavities tested in a dedicated cavity pair cryomodule at 2K. The safety margin to the BBU threshold budget at 12 GeV has been assessed.

  19. Flammable gas tank safety program: Technical basis for gas analysis and monitoring

    International Nuclear Information System (INIS)

    Estey, S.D.

    1998-01-01

    Several Hanford waste tanks have been observed to exhibit periodic releases of significant quantities of flammable gases. Because potential safety issues have been identified with this type of waste behavior, applicable tanks were equipped with instrumentation offering the capability to continuously monitor gases released from them. This document was written to cover three primary areas: (1) describe the current technical basis for requiring flammable gas monitoring, (2) update the technical basis to include knowledge gained from monitoring the tanks over the last three years, (3) provide the criteria for removal of Standard Hydrogen Monitoring System(s) (SHMS) from a waste tank or termination of other flammable gas monitoring activities in the Hanford Tank farms

  20. Upgrade of Common Cause Failure Modelling of NPP Krsko PSA

    International Nuclear Information System (INIS)

    Vukovic, I.; Mikulicic, V.; Vrbanic, I.

    2006-01-01

    Over the last thirty years the probabilistic safety assessments (PSA) have been increasingly applied in technical engineering practice. Various failure modes of system of concern are mathematically and explicitly modelled by means of fault tree structure. Statistical independence of basic events from which the fault tree is built is not acceptable for an event category referred to as common cause failures (CCF). Based on overview of current international status of modelling of common cause failures in PSA several steps were made related to primary technical basis for methodology and data used for CCF model upgrade project in NPP Krsko (NEK) PSA. As a primary technical basis for methodological aspects of CCF modelling in Krsko PSA the following documents were considered: NUREG/CR-5485, NUREG/CR-4780, and Westinghouse Owners Group documents (WOG) WCAP-15674 and WCAP-15167. Use of these documents is supported by the most relevant guidelines and standards in the field, such as ASME PRA Standard and NRC Regulatory Guide 1.200. WCAP documents are in compliance with NUREG/CR-5485 and NUREG/CR-4780. Additionally, they provide WOG perspective on CCF modelling, which is important to consider since NEK follows WOG practice in resolving many generic and regulatory issues. It is, therefore, desirable that NEK CCF methodology and modelling is in general accordance with recommended WOG approaches. As a primary basis for CCF data needed to estimate CCF model parameters and their uncertainty, the main used documents were: NUREG/CR-5497, NUREG/CR-6268, WCAP-15167, and WCAP-16187. Use of NUREG/CR-5497 and NUREG/CR-6268 as a source of data for CCF parameter estimating is supported by the most relevant industry and regulatory PSA guides and standards currently existing in the field, including WOG. However, the WCAP document WCAP-16187 has provided a basis for CCF parameter values specific to Westinghouse PWR plants. Many of events from NRC / INEEL database were re-classified in WCAP

  1. Fusion of eastern and western technology in VVER 1000 NPP upgrade

    International Nuclear Information System (INIS)

    Ubra, O.; Fleischhans, J.; Kveton, M.

    1997-01-01

    An extensive modernization program upgrading two units of VVER 1000 type of the Czech nuclear power plant (NPP) Temelin to meet the latest international standards is presented. The program is based primarily on combination of eastern and western technology and it has been implemented during plant construction. The NPP Temelin was originally designed according to the standards of the former Soviet Union. After a series of reviews in the 1990s, a decision was made by the Temelin management of upgrade the design of the plant, including the supply of fuel and instrumentation and control system by a western company. The adoption of western technology and practices has helped to solve a large number of IAEA safety issues related to design and operation of VVER 1000 NPP. Details on the current Temelin design and other related safety matters are presented

  2. Safety of reactors built according to earlier standards (WWER 440/V230 type)

    International Nuclear Information System (INIS)

    Misak, J.; Rohar, S.

    1995-01-01

    The problems of safety of WWER-440/V-230 type reactors are discussed, and the following conclusions are made. (1) The reactors have a very good operational record. (2) The reactors have serious design shortcomings, which should be eliminated by safety upgrading. Core damage frequency should be further reduced. (3) PSA methods constitute an appropriate tool for assessment of plant vulnerability to some initiating events and malfunctions, for prioritization of upgrading measures and for tolerability of deviations from current safety standards. (4) The most important safety merits, such as a large thermal inertia and low rupture probability, should be properly taken into account in the analysis. (5) Extensive safety upgrading is feasible and can lead to a considerable risk reduction. In certain circumstances such upgrading is the least expensive option even though the total cost is much higher than the initial plant construction cost. (6) Properly upgraded, the reactor units may be operable until better power resources are available within the country. (7) The existing gap between the technological and political judgements of nuclear safety should be reduced continuously by information exchange improvements. (8) A unified approach to nuclear safety should be adopted for all nuclear reactors (not just WWERs) built to earlier standards. 5 tabs., 1 fig

  3. Analysis of Reactor Pressurized Thermal Shock Conditions Considering Upgrading of Systems Important to Safety

    International Nuclear Information System (INIS)

    Mazurok, A.S; Vyshemirskyij, M.P.

    2015-01-01

    The paper analyzes conditions of pressurized thermal shock on the reactor pressure vessel taking into account upgrading of the emergency core cooling system and primary overpressure protection system. For representative accident scenarios, calculation and comparative analysis was carried out. These scenarios include a small leak from the hot leg and PRZ SV stuck opening with re closure after 3600 sec and 3 SG heat transfer tube rupture. The efficiency of mass flow control by valves on the pump head (emergency core cooling systems) and cold overpressure protection (primary overpressure protection system) was analyzed. The thermal hydraulic model for RELAP5/Mod3.2 code with detailed downcomer (DC) model and changes in accordance with upgrades was used for calculations. Detailed (realistic) modeling of piping and equipment was performed. The upgrades prevent excessive primary cooling and, consequently, help to preserve the RPV integrity and to avoid the formation of a through crack, which can lead to a severe accident

  4. Optimization of the nuclear power engineering safety on the basis of social and economic parameters

    International Nuclear Information System (INIS)

    Kozlov, V.F.; Kuz'min, I.I.; Lystsov, V.N.; Amosova, T.V.; Makhutov, N.A.; Men'shikov, V.F.

    1995-01-01

    Principle of optimization of nuclear power engineering safety is presented on the basis of estimating the risks to the man's health with an account of peculiarities of socio-economic system and other types of economic activities in the region. Average expected duration of forthcoming life and costs of its prolongation serve as a unit for measuring the man's safety. It is shown that if the expenditures on NPP technical safety exceed the scientifically substantiated costs for this region with application of the above principle, than the risk for population will exceed the minimum achievable level. 8 refs., 2 figs., 1 tab

  5. To the problem of the statistical basis of evaluation of the mechanical safety factor

    International Nuclear Information System (INIS)

    Tsyganov, S.V.

    2009-01-01

    The methodology applied for the safety factor assessment of the WWER fuel cycles uses methods and terms of statistics. Value of the factor is calculated on the basis of estimation of probability to meet predefined limits. Such approach demands the special attention to the statistical properties of parameters of interest. Considering the mechanical constituents of the engineering factor it is assumed uncertainty factors of safety parameters are stochastic values. It characterized by probabilistic distributions that can be unknown. Traditionally in the safety factor assessment process the unknown parameters are estimated from the conservative points of view. This paper analyses how the refinement of the factors distribution parameters is important for the assessment of the mechanical safety factor. For the analysis the statistical approach is applied for modelling of different type of factor probabilistic distributions. It is shown the significant influence of the shape and parameters of distributions for some factors on the value of mechanical safety factor. (Authors)

  6. To the problem of the statistical basis of evaluation of the mechanical safety factor

    International Nuclear Information System (INIS)

    Tsyganov, S.

    2009-01-01

    The methodology applied for the safety factor assessment of the VVER fuel cycles uses methods and terms of statistics. Value of the factor is calculated on the basis of estimation of probability to meet predefined limits. Such approach demands the special attention to the statistical properties of parameters of interest. Considering the mechanical constituents of the engineering factor it is assumed uncertainty factors of safety parameters are stochastic values. It characterized by probabilistic distributions that can be unknown. Traditionally in the safety factor assessment process the unknown parameters are estimated from the conservative points of view. This paper analyses how the refinement of the factors distribution parameters is important for the assessment of the mechanical safety factor. For the analysis the statistical approach is applied for modelling of different type of factor probabilistic distributions. It is shown the significant influence of the shape and parameters of distributions for some factors on the value of mechanical safety factor. (author)

  7. UPGRADES

    CERN Document Server

    D. Contardo and J. Spalding

    2013-01-01

    There is very good progress in the execution of the LS1 projects and in launching construction of the Phase 1 upgrades. We focus here on two main achievements since the last CMS Week. The approval of the third Phase 1 TDR The preparation of the L1 Trigger Upgrade Technical Design Report has been a major effort of the collaboration at the beginning of this year, especially to develop supporting Trigger menu and physics performance studies. These studies have demonstrated the efficiency of the upgraded system to ensure low lepton and jet trigger thresholds, leading to a significant increase of the acceptance for the Higgs measurements, in the associated production mode and in the ττ decays, as well as for the stop searches involving multiple jets in the final state. The TDR was submitted to the LHCC in May and approved at the June committee meeting. It is now a public document, completing the series of the three TDRs describing the Phase 1 upgrades, with the new Pixel system and the HCAL rea...

  8. FPGAs Emulate Microprocessors-A Successful Case for HFC NPP Digital I and C Upgrade

    International Nuclear Information System (INIS)

    Hsu, Allen; Crow, Ivan; Reese, Carl; Kim, Jong; Yang, Steve

    2014-01-01

    Field Programmable Gate Arrays (FPGAs), as programmable logic devices (PLDs) have gained a great deal of interests for implementing safety I and C applications in nuclear power plants (NPPs) largely owing to the FPGAs'potential advantage over the currently more common microprocessor-based digital I and C applications. First of all, FPGAs have adequate capabilities for most digital I and C applications in NPPs. Secondly, FPGAs provide products with longer lifetime, improve testability, and reduce the drift which occurs in analog-based systems, from hardware perspective. Thirdly, FPGAs, from software perspective, can be made simpler, less reliant on complex software such as operating systems, which should make FPGAs easier to qualify for nuclear safety applications. Fourthly, FPGAs are less vulnerable to cyber attacks when FPGAs implement the I and C systems that do not contain high-level, general purpose software that may be easily subjected to malicious modifications. Finally, FPGAs can bring cost reduction in an I and C digital upgrade because FPGAs can provide simpler licensing process than microprocessor-based digital I and C, and FPGAs can be implemented more efficiently. This paper will present one successful case for YGN Unit I and C upgrade using FPGA-based components to replace the obsolete Intel 8085 Microprocessor-based controllers. In this case, FPGAs emulated the process of the existing microprocessors and interpreted the execution of CPU processing. More than 160 of the FPGA-based SBC-01 controllers replacing the Intel 8085 Microprocessor-based Printed Circuit Boards have been installed and running successfully for safety I and C applications over the last five years. In this upgrade, the new FPGA-based controller board SBC-01 emulated the functions of Intel 8085 microprocessor correctly. It is a successful and cost-effective upgrade.vIn this paper, lifecycle design and implementation process and rigorous V and V activities that were used in the

  9. FPGAs Emulate Microprocessors-A Successful Case for HFC NPP Digital I and C Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, Allen; Crow, Ivan; Reese, Carl; Kim, Jong; Yang, Steve [Doosan HF Controls Corp, Carrollton (United States)

    2014-08-15

    Field Programmable Gate Arrays (FPGAs), as programmable logic devices (PLDs) have gained a great deal of interests for implementing safety I and C applications in nuclear power plants (NPPs) largely owing to the FPGAs'potential advantage over the currently more common microprocessor-based digital I and C applications. First of all, FPGAs have adequate capabilities for most digital I and C applications in NPPs. Secondly, FPGAs provide products with longer lifetime, improve testability, and reduce the drift which occurs in analog-based systems, from hardware perspective. Thirdly, FPGAs, from software perspective, can be made simpler, less reliant on complex software such as operating systems, which should make FPGAs easier to qualify for nuclear safety applications. Fourthly, FPGAs are less vulnerable to cyber attacks when FPGAs implement the I and C systems that do not contain high-level, general purpose software that may be easily subjected to malicious modifications. Finally, FPGAs can bring cost reduction in an I and C digital upgrade because FPGAs can provide simpler licensing process than microprocessor-based digital I and C, and FPGAs can be implemented more efficiently. This paper will present one successful case for YGN Unit I and C upgrade using FPGA-based components to replace the obsolete Intel 8085 Microprocessor-based controllers. In this case, FPGAs emulated the process of the existing microprocessors and interpreted the execution of CPU processing. More than 160 of the FPGA-based SBC-01 controllers replacing the Intel 8085 Microprocessor-based Printed Circuit Boards have been installed and running successfully for safety I and C applications over the last five years. In this upgrade, the new FPGA-based controller board SBC-01 emulated the functions of Intel 8085 microprocessor correctly. It is a successful and cost-effective upgrade.vIn this paper, lifecycle design and implementation process and rigorous V and V activities that were used in the

  10. Safety in Swiss nuclear power plants

    International Nuclear Information System (INIS)

    Cederqvist, H.

    1992-01-01

    Safety-related facilities and equipment are continuously backfitted in Swiss nuclear power plants. In the Beznau-1 and -2 nuclear generating units, the measures taken under the heading of 'Backfitting of Emergency Systems' included provisions to enhance the protection against earthquakes, airplane crash, and fire; in addition, the emergency power system was upgraded. In Muehleberg, the stack exhaust air monitoring system was optimized. The containment pressure suppression system of the plant has been designed to withstand a hypothetical accident exceeding the design basis. The BKM-Crud computer simulation model simulates steps taken to reduce radiation exposure. The power of Swiss nuclear power stations will be raised by 4% to 15% within the 'Energy 2000' action program. (orig.) [de

  11. Retrofitting a spent fuel pool spray system for alternative cooling as a strategy for beyond design basis events

    Energy Technology Data Exchange (ETDEWEB)

    Hartmann, Christoph; Vujic, Zoran [Westinghouse Electric Germany GmbH, Mannheim (Germany)

    2017-06-15

    Due to requirements for nuclear power plants to withstand beyond design basis accidents, including events such as happened in 2011 in the Fukushima Daiichi Nuclear Power Plant in Japan, alternative cooling of spent fuel is needed. Alternative spent fuel cooling can be provided by a retrofitted spent fuel pool spray system based on the AP1000 plant design. As part of Krsko Nuclear Power Plant's Safety Upgrade Program, Krsko Nuclear Power Plant decided on, and Westinghouse successfully designed a retrofit of the AP1000 {sup registered} plant spent fuel pool spray system to provide alternative spent fuel cooling.

  12. Conceptual design of a neutron camera for MAST Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Weiszflog, M., E-mail: matthias.weiszflog@physics.uu.se; Sangaroon, S.; Cecconello, M.; Conroy, S.; Ericsson, G.; Klimek, I. [Department of Physics and Astronomy, Uppsala University, EURATOM-VR Association, Uppsala (Sweden); Keeling, D.; Martin, R. [CCFE, Culham Science Centre, Abingdon (United Kingdom); Turnyanskiy, M. [ITER Physics Department, EFDA CSU Garching, Boltzmannstrae 2, D-85748 Garching (Germany)

    2014-11-15

    This paper presents two different conceptual designs of neutron cameras for Mega Ampere Spherical Tokamak (MAST) Upgrade. The first one consists of two horizontal cameras, one equatorial and one vertically down-shifted by 65 cm. The second design, viewing the plasma in a poloidal section, also consists of two cameras, one radial and the other one with a diagonal view. Design parameters for the different cameras were selected on the basis of neutron transport calculations and on a set of target measurement requirements taking into account the predicted neutron emissivities in the different MAST Upgrade operating scenarios. Based on a comparison of the cameras’ profile resolving power, the horizontal cameras are suggested as the best option.

  13. MAST Upgrade – Construction Status

    Energy Technology Data Exchange (ETDEWEB)

    Milnes, Joe, E-mail: Joe.Milnes@ccfe.ac.uk; Ayed, Nizar Ben; Dhalla, Fahim; Fishpool, Geoff; Hill, John; Katramados, Ioannis; Martin, Richard; Naylor, Graham; O’Gorman, Tom; Scannell, Rory

    2015-10-15

    Highlights: • Outlines unique capability of MAST-U, including divertor and diagnostic capability. • Describes progress made in the manufacture and assembly of key MAST-U components. • Highlights the design challenges that have been overcome. • Lists the key lessons learned thus far in the project. - Abstract: The Mega Amp Spherical Tokamak (MAST) is the centre piece of the UK fusion research programme. In 2010, a MAST Upgrade programme was initiated with three primary objectives, to contribute to: (1) testing reactor concepts (in particular exhaust solutions via a flexible divertor allowing Super-X and other extended leg configurations); (2) adding to the knowledge base for ITER (by addressing important plasma physics questions and developing predictive models to help optimise ITER performance of ITER) and (3) exploring the feasibility of using a spherical tokamak as the basis for a fusion Component Test Facility. With the project mid-way through its construction phase, progress will be reported on a number of the critical subsystems. This will include manufacture and assembly of the coils, armour and support structures that make up the new divertors, construction of the new set coils that make up the centre column, installation of the new power supplies for powering the divertor coils and enhanced TF coil set, progress in delivering the upgraded diagnostic capability, the modification and upgrading of the NBI heating systems and the complete overhaul of the machine control infrastructure, including a new control room with full remote participation facilities.

  14. ITER safety task NID-5a: ITER tritium environmental source terms - safety analysis basis

    International Nuclear Information System (INIS)

    Natalizio, A.; Kalyanam, K.M.

    1994-09-01

    The Canadian Fusion Fuels Technology Project's (CFFTP) is part of the contribution to ITER task NID-5a, Initial Tritium Source Term. This safety analysis basis constitutes the first part of the work for establishing tritium source terms and is intended to solicit comments and obtain agreement. The analysis objective is to provide an early estimate of tritium environmental source terms for the events to be analyzed. Events that would result in the loss of tritium are: a Loss of Coolant Accident (LOCA), a vacuum vessel boundary breach. a torus exhaust line failure, a fuelling machine process boundary failure, a fuel processing system process boundary failure, a water detritiation system process boundary failure and an isotope separation system process boundary failure. 9 figs

  15. Physics design of the upgraded TREAT reactor

    International Nuclear Information System (INIS)

    Bhattacharyya, S.K.; Lell, R.M.; Liaw, J.R.; Ulrich, A.J.; Wade, D.C.; Yang, S.T.

    1980-01-01

    With the deferral of the Safety Test Facility (STF), the TREAT Upgrade (TU) reactor has assumed a lead role in the US LMFBR safety test program for the foreseeable future. The functional requirements on TU require a significant enhancement of the capability of the current TREAT reactor. A design of the TU reactor has been developed that modifies the central 11 x 11 fuel assembly array of the TREAT reactor such as to provide the increased source of hard spectrum neutrons necessary to meet the functional requirements. A safety consequence of the increased demands on TU is that the self limiting operation capability of TREAT has proved unattainable, and reliance on a safety grade Plant Protection System is necessary to ensure that no clad damage occurs under postulated low-probability reactivity accidents. With that constraint, the physics design of TU provides a means of meeting the functional requirements with a high degree of confidence

  16. Overview of Mobile Equipment Used in Case of Beyond Design Basis Accident at NPP Krsko

    International Nuclear Information System (INIS)

    Lukacevic, H.; Kopinc, D.; Ivanjko, M.

    2016-01-01

    Terrorist attack in USA in the September 11, 2001 and accident at the Fukushima - Daiichi Nuclear Power Station in the March 11, 2011 highlight the importance of mitigating strategies in responding to Beyond Design Basis Accident (BDBA), while ensuring cooling of reactor core, containment and spent fuel pool. Nuclear Power Plant Krsko (NEK) has acquired additional mobile equipment and made necessary modifications on existing systems for the connection of this equipment (fast couplers). Usage of mobile equipment is not only limited to design basis accident (DBA), but, also to prevent and mitigate the consequences in case of BDBA, when other plant systems are not available. NEK also decided to take steps for upgrade of safety measures and prepared Safety Upgrade Program (SUP), which is consistent with the nuclear industry response to the Fukushima accident and is implementing main projects and modifications related to SUP. NEK mobile equipment is not required to operate under normal reactor plant operation except for periodic surveillance testing and is incorporated into the normal training process. Equipment is dislocated from the reactor building and most of the equipment is located in the new building, able to withstand extreme natural events, including earthquakes and tornadoes. The usage of all mobile equipment is prescribed as an additional option in NEK operating procedures in following cases and enables following options: filling various tanks, filling the steam generators, filling the containment, additional compressed air source, spent fuel pool refilling and spraying, alternative power supply. This document provides an overview of NEK mobile equipment, which consists of various mobile fire protection pumps, air compressors, protective equipment, fire trucks, diesel generators. Sufficient fuel supply for the equipment is provided on site for a minimum three days of operation. (author).

  17. Safety study application guide

    International Nuclear Information System (INIS)

    1993-07-01

    Martin Marietta Energy Systems, Inc., (Energy Systems) is committed to performing and documenting safety analyses for facilities it manages for the Department of Energy (DOE). Included are analyses of existing facilities done under the aegis of the Safety Analysis Report Upgrade Program, and analyses of new and modified facilities. A graded approach is used wherein the level of analysis and documentation for each facility is commensurate with the magnitude of the hazard(s), the complexity of the facility and the stage of the facility life cycle. Safety analysis reports (SARs) for hazard Category 1 and 2 facilities are usually detailed and extensive because these categories are associated with public health and safety risk. SARs for Category 3 are normally much less extensive because the risk to public health and safety is slight. At Energy Systems, safety studies are the name given to SARs for Category 3 (formerly open-quotes lowclose quotes) facilities. Safety studies are the appropriate instrument when on-site risks are limited to irreversible consequences to a few people, and off-site consequences are limited to reversible consequences to a few people. This application guide provides detailed instructions for performing safety studies that meet the requirements of DOE Orders 5480.22, open-quotes Technical Safety Requirements,close quotes and 5480.23, open-quotes Nuclear Safety Analysis Reports.close quotes A seven-chapter format has been adopted for safety studies. This format allows for discussion of all the items required by DOE Order 5480.23 and for the discussions to be readily traceable to the listing in the order. The chapter titles are: (1) Introduction and Summary, (2) Site, (3) Facility Description, (4) Safety Basis, (5) Hazardous Material Management, (6) Management, Organization, and Institutional Safety Provisions, and (7) Accident Analysis

  18. Analysis and development of the automated emergency algorithm to control primary to secondary LOCA for SUNPP safety upgrading

    International Nuclear Information System (INIS)

    Kim, V.; Kuznetsov, V.; Balakan, G.; Gromov, G.; Krushynsky, A.; Sholomitsky, S.; Lola, I.

    2007-01-01

    The paper presents the results of the study conducted to support planned modernization of the South Ukraine nuclear power plant. The objective of the analysis has been to develop the automated emergency control algorithm for primary to secondary LOCA accident for SUNPP WWER-1000 safety upgrading. According to the analyses performed in the framework of safety assesment report, given accident is the most complex for control and has the largest contribution into the core damage frequency value. This is because of initial event diagnostics is difficult, emergency control is complicated for personnel, time available for decision making and actions performing is limited with coolant inventory for make-up, probability of steam dump valves on affected steam generator non-closing after opening is high, and as a consequence containment bypass, irretrievable loss of coolant and radioactive materials release into the environment are possible. Unit design modifications are directed on expansion of safety systems capabilities to overcome given accident and to facilitate the personnel actions on emergency control. Safety systems modification according to developed algorithm will allow to simplify accident control by personnel and enable to control the ECCS discharge limiting pressure below the affected steam generator steam dump valve opening pressure, and decrease the probability of the containment bypass sequences. The analysis of the primary-to-secondary LOCA thermal-hydraulics has been conducted with RELAP5/Mod 3.2, and involved development of the dedicated analytical model, calculations of various plant response accident scenarios, conducting of plant personnel intervention analyses using full-scale simulator, development and justification of the emergency control algorithm aimed on the minimization of negative consequences of the primary-to-secondary LOCA (Authors)

  19. Review of the treat upgrade reactor scram system reliability analysis

    International Nuclear Information System (INIS)

    Montague, D.F.; Fussell, J.B.; Krois, P.A.; Morelock, T.C.; Knee, H.E.; Manning, J.J.; Haas, P.M.; West, K.W.

    1984-10-01

    In order to resolve some key LMFBR safety issues, ANL personnel are modifying the TREAT reactor to handle much larger experiments. As a result of these modifications, the upgraded Treat reactor will not always operate in a self-limited mode. During certain experiments in the upgraded TREAT reactor, it is possible that the fuel could be damaged by overheating if, once the computer systems fail, the reactor scram system (RSS) fails on demand. To help ensure that the upgraded TREAT reactor is shut down when required, ANL personnel have designed a triply redundant RSS for the facility. The RSS is designed to meet three reliability goals: (1) a loss of capability failure probability of 10 -9 /demand (independent failures only); (2) an inadvertent shutdown probability of 10 -3 /experiment; and (3) protection agaist any known potential common cause failures. According to ANL's reliability analysis of the RSS, this system substantially meets these goals

  20. Setting clear expectations for safety basis development

    International Nuclear Information System (INIS)

    MORENO, M.R.

    2003-01-01

    DOE-RL has set clear expectations for a cost-effective approach for achieving compliance with the Nuclear Safety Management requirements (10 CFR 830, Nuclear Safety Rule) which will ensure long-term benefit to Hanford. To facilitate implementation of these expectations, tools were developed to streamline and standardize safety analysis and safety document development resulting in a shorter and more predictable DOE approval cycle. A Hanford Safety Analysis and Risk Assessment Handbook (SARAH) was issued to standardized methodologies for development of safety analyses. A Microsoft Excel spreadsheet (RADIDOSE) was issued for the evaluation of radiological consequences for accident scenarios often postulated for Hanford. A standard Site Documented Safety Analysis (DSA) detailing the safety management programs was issued for use as a means of compliance with a majority of 3009 Standard chapters. An in-process review was developed between DOE and the Contractor to facilitate DOE approval and provide early course correction. As a result of setting expectations and providing safety analysis tools, the four Hanford Site waste management nuclear facilities were able to integrate into one Master Waste Management Documented Safety Analysis (WM-DSA)

  1. Radiation safety experience in upgrading 2-5% heavy water wastes at Heavy Water Plant, Nangal (Preprint No. SA-7)

    International Nuclear Information System (INIS)

    Sadhukhan, H.K.; Behl, D.; Ramraj; Iyengar, T.S.; Sadarangani, S.H.; Vaze, P.K.; Soman, S.D.

    1989-04-01

    This paper describes the radiological safety experience in upgrading 2-5% heavy water wastes at Heavy Water Plant at Nangal at the third stage electrolysers. The feed water concentrations at the third stage electrolyer was determined after a safety analysis study and pilot plant experiment, which gave the optimal concentrations of 1 to 1.5 mCi (3.7 to 5.5 x 10 7 Bq) per litre per minute feed from a submerged SS tank containing 2-5% heavy water wastes. This process not only yielded an efficient recovery of reactor grade heavy water but contained the tritium activity in the third stage electrolysers and in the final product viz., heavy water. The tritium concentrations were continuously monitore d by liquid scintillation counting method at all the three stages of electrolysis plant, the distillation plant, the heavy water filling rooms, the drains, the ambient air, the product fertilizer (calcium ammonia nitrate) and the Sutlej River and found to be well within the safety limits set for general public at large. The HD and D 2 process streams in the palnt were monitored using fill-in type of ionization chambers designed for the purpose, which served a D 2 inventory check as well. There was no internal exposure to any personnel during the entire period of programme. (author). 2 tabs

  2. Primary Vertex Reconstruction for Upgrade at LHCb

    CERN Document Server

    Wanczyk, Joanna

    2016-01-01

    The aim of the LHCb experiment is the study of beauty and charm hadron decays with the main focus on CP violating phenomena and searches for physics beyond the Standard Model through rare decays. At the present, the second data taking period is ongoing, which is called Run II. After 2018 during the long shutdown, the replacement of signicant parts of the LHCb detector is planned. One of main changes is upgrade of the present software and hardware trigger to a more rapid full software trigger. Primary Vertex (PV) is a basis for the further tracking and it is sensitive to the LHC running conditions, which are going to change for the Upgrade. In particular, the center-of-mass collision energy should reach the maximum value of 14 TeV. As a result the quality of the reconstruction has to be studied and the reconstruction algorithms have to be optimized.

  3. Tank farm instrumentation and data acquisition/management upgrade plan

    International Nuclear Information System (INIS)

    Scaief, C.C. III.

    1994-01-01

    This plan provides the strategy, implementation, and schedule for upgrading tank farm instrumentation, data acquisition and data management. The focus is on surveillance parameters to verify and maintain tank safety. The criteria do not necessarily constitute mandatory requirements but are based upon engineering judgement and best available information. Schedules reflect preliminary funding for FY95. For out years they are best engineering judgment

  4. Fire protection upgrading of four Russian 440/230 VVER units

    International Nuclear Information System (INIS)

    Corsini, G.; Yelfimov, S.

    1995-01-01

    The main goal of TACIS 3.6 a project funded by the Commission of the European Communities (CEC), was the front-end engineering for upgrading the Fire Protection System (FPS) of the safety-related equipment of Novovoronezh, Units 3 and 4, and Kola, Units 1 and 2, VVER 440/230 nuclear power plants. As a first step, all the safety-related equipment had to be identified, evaluation criteria had to be established and the existing FPS reviewed against the criteria. In the second step, the selection of the upgrading measures, depending on feasibility and cost estimate, has been accomplished, room by room. The third step, carried out on schedule and completed end July 95, has been essentially the preparation of the Technical Specifications for procurement of the needed equipment including remaining detail engineering. The Russian sub-contactor Atom Energo Project (AEP), who have been the designers of these older NPP s, have done the work with the Italian Ansaldo as the consultants of their Russian colleagues. Practical aspects of the engineering work are discussed and examples of improvements selected for retrofitting described. (author)

  5. Silicon vertex detector upgrade in the ALPHA experiment

    CERN Document Server

    Amole, C; Ashkezari, M.D; Baquero-Ruiz, M; Bertsche, W; Burrows, C; Butler, E; Capra, A; Cesar, C.L; Chapman, S; Charlton, M; Deller, A; Eriksson, S; Fajans, J; Friesen, T; Fujiwara, M.C; Gill, D.R; Gutierrez, A; Hangst, J.S; Hardy, W.N; Hayden, M.E; Humphries, A.J; Isaac, C.A; Jonsell, S; Kurchaninov, L; Little, A; Madsen, N; McKenna, J.T.K; Menary, S; Napoli, S.C; Nolan, P; Olchanski, K; Olin, A; Povilus, A; Pusa, P; Rasmussen, C.Ø; Robicheaux, F; Sacramento, R.L; Sampson, J.A; Sarid, E; Seddon, D; Silveira, D.M; So, C; Stracka, S; Tharp, T; Thompson, R.I; Thornhill, J; Tooley, M.P; Van Der Werf, D.P; Wells, D

    2013-01-01

    The Silicon Vertex Detector (SVD) is the main diagnostic tool in the ALPHA-experiment. It provides precise spatial and timing information of antiproton (antihydrogen) annihilation events (vertices), and most importantly, the SVD is capable of directly identifying and analysing single annihilation events, thereby forming the basis of ALPHA ' s analysis. This paper describes the ALPHA SVD and its upgrade, installed in the ALPHA ' s new neutral atom trap.

  6. Silicon vertex detector upgrade in the ALPHA experiment

    Energy Technology Data Exchange (ETDEWEB)

    Amole, C. [Department of Physics and Astronomy, York University, Toronto, ON, M3J 1P3 (Canada); Andresen, G.B. [Department of Physics and Astronomy, Aarhus University, DK-8000 Aarhus C (Denmark); Ashkezari, M.D. [Department of Physics, Simon Fraser University, Burnaby, BC, V5A 1S6 (Canada); Baquero-Ruiz, M. [Department of Physics, University of California at Berkeley, Berkeley, CA 94720-7300 (United States); Bertsche, W. [School of Physics and Astronomy, University of Manchester, M13 9PL Manchester (United Kingdom); The Cockcroft Institute, Daresbury Laboratory, WA4 4AD Warrington (United Kingdom); Burrows, C. [Department of Physics, College of Science, Swansea University, Swansea SA2 8PP (United Kingdom); Butler, E. [Physics Department, CERN, CH-1211 Geneva 23 (Switzerland); Capra, A. [Department of Physics and Astronomy, York University, Toronto, ON, M3J 1P3 (Canada); Cesar, C.L. [Instituto de Física, Universidade Federal do Rio de Janeiro, Rio de Janeiro 21941-972 (Brazil); Chapman, S. [Department of Physics, University of California at Berkeley, Berkeley, CA 94720-7300 (United States); Charlton, M.; Deller, A.; Eriksson, S. [Department of Physics, College of Science, Swansea University, Swansea SA2 8PP (United Kingdom); Fajans, J. [Department of Physics, University of California at Berkeley, Berkeley, CA 94720-7300 (United States); Lawrence Berkeley National Laboratory, Berkeley, CA 94720 (United States); Friesen, T. [Department of Physics and Astronomy, University of Calgary, Calgary, Alberta T2N 1N4 (Canada); Fujiwara, M.C. [Department of Physics and Astronomy, University of Calgary, Calgary, Alberta T2N 1N4 (Canada); TRIUMF, 4004 Wesbrook Mall, Vancouver, BC, V6T 2A3 (Canada); Gill, D.R. [TRIUMF, 4004 Wesbrook Mall, Vancouver, BC, V6T 2A3 (Canada); Gutierrez, A. [Department of Physics and Astronomy, University of British Columbia, Vancouver, BC, V6T 1Z4 (Canada); and others

    2013-12-21

    The Silicon Vertex Detector (SVD) is the main diagnostic tool in the ALPHA-experiment. It provides precise spatial and timing information of antiproton (antihydrogen) annihilation events (vertices), and most importantly, the SVD is capable of directly identifying and analysing single annihilation events, thereby forming the basis of ALPHA's analysis. This paper describes the ALPHA SVD and its upgrade, installed in the ALPHA's new neutral atom trap.

  7. Technical basis for environmental qualification of microprocessor-based safety-related equipment in nuclear power plants

    International Nuclear Information System (INIS)

    Korsah, K.; Wood, R.T.; Hassan, M.; Tanaka, T.J.

    1998-01-01

    This document presents the results of studies sponsored by the Nuclear Regulatory Commission (NRC) to provide the technical basis for environmental qualification of computer-based safety equipment in nuclear power plants. The studies were conducted by Oak Ridge National Laboratory (ORNL), Sandia National Laboratories (SNL), and Brookhaven National Laboratory (BNL). The studies address the following: (1) adequacy of the present test methods for qualification of digital I and C systems; (2) preferred (i.e., Regulatory Guide-endorsed) standards; (3) recommended stressors to be included in the qualification process during type testing; (4) resolution of need for accelerated aging for equipment to be located in a benign environment; and (5) determination of an appropriate approach for addressing the impact of smoke in digital equipment qualification programs. Significant findings from the studies form the technical basis for a recommended approach to the environmental qualification of microprocessor-based safety-related equipment in nuclear power plants

  8. Technical basis for environmental qualification of microprocessor-based safety-related equipment in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Korsah, K.; Wood, R.T. [Oak Ridge National Lab., TN (United States); Hassan, M. [Brookhaven National Lab., Upton, NY (United States); Tanaka, T.J. [Sandia National Labs., Albuquerque, NM (United States)

    1998-01-01

    This document presents the results of studies sponsored by the Nuclear Regulatory Commission (NRC) to provide the technical basis for environmental qualification of computer-based safety equipment in nuclear power plants. The studies were conducted by Oak Ridge National Laboratory (ORNL), Sandia National Laboratories (SNL), and Brookhaven National Laboratory (BNL). The studies address the following: (1) adequacy of the present test methods for qualification of digital I and C systems; (2) preferred (i.e., Regulatory Guide-endorsed) standards; (3) recommended stressors to be included in the qualification process during type testing; (4) resolution of need for accelerated aging for equipment to be located in a benign environment; and (5) determination of an appropriate approach for addressing the impact of smoke in digital equipment qualification programs. Significant findings from the studies form the technical basis for a recommended approach to the environmental qualification of microprocessor-based safety-related equipment in nuclear power plants.

  9. Report of the Fukushima nuclear accident by the National Academy of Science. Lessons learned from the Fukushima nuclear accident for improving safety of U.S. nuclear plants

    International Nuclear Information System (INIS)

    Nariai, Hideki

    2014-01-01

    U.S. National Academy of Science investigated the accident at the Fukushima Daiichi nuclear plant initiated by the Great East Japan Earthquake for two years and published a draft report in July 24, 2014. Investigation results were summarized in nine new findings and made ten recommendations in a wide horizon; (1) hardware countermeasures against severe accidents and training of operators, (2) upgrade of risk assessment capability for beyond design basis accident, (3) incorporation of new information about hazards in safety regulations, (4) needed improvement of off-site emergency preparedness, and (5) improvements of nuclear safety culture. New information about hazards related with tsunami assessment, new risk assessment for beyond design basis accident, advice of foreigner resident evacuations, regulatory capture, and safety culture and regulator's specialty were discussed as Japanese issues. (T. Tanaka)

  10. Refurbishment and safety up-gradation of Cirus Reactor

    International Nuclear Information System (INIS)

    Rao, D.V.H.

    2004-01-01

    Cirus, a 40 MWth, vertical tank type research reactor, having a wide range of research facilities, was commissioned in 1960. This research facility has been operated and utilized extensively for nearly four decades. With a view to assess the residual life of the reactor, detailed ageing studies were carried out. Based on this, refurbishment work for life extension was undertaken. During this work, additional safety features were incorporated to improve the overall safety of the reactor. This lecture details the methodologies used for ageing studies and refurbishment activities for life extension with enhanced safety. (author)

  11. Rationale and design of the BUDAPEST-CRT Upgrade Study: a prospective, randomized, multicentre clinical trial.

    Science.gov (United States)

    Merkely, Bela; Kosztin, Annamaria; Roka, Attila; Geller, Laszlo; Zima, Endre; Kovacs, Attila; Boros, Andras Mihaly; Klein, Helmut; Wranicz, Jerzy K; Hindricks, Gerhard; Clemens, Marcell; Duray, Gabor Z; Moss, Arthur J; Goldenberg, Ilan; Kutyifa, Valentina

    2017-09-01

    There is lack of conclusive evidence from randomized clinical trials on the efficacy and safety of upgrade to cardiac resynchronization therapy (CRT) in patients with implanted pacemakers (PM) or defibrillators (ICD) with reduced left ventricular ejection fraction (LVEF) and chronic heart failure (HF). The BUDAPEST-CRT Upgrade Study was designed to compare the efficacy and safety of CRT upgrade from conventional PM or ICD therapy in patients with intermittent or permanent right ventricular (RV) septal/apical pacing, reduced LVEF, and symptomatic HF. The BUDAPEST-CRT study is a prospective, randomized, multicentre, investigator-sponsored clinical trial. A total of 360 subjects will be enrolled with LVEF ≤ 35%, NYHA functional classes II-IVa, paced QRS ≥ 150 ms, and a RV pacing ≥ 20%. Patients will be followed for 12 months. Randomization is performed in a 3:2 ratio (CRT-D vs. ICD). The primary composite endpoint is all-cause mortality, a first HF event, or less than 15% reduction in left ventricular (LV) end-systolic volume at 12 months. Secondary endpoints are all-cause mortality, all-cause mortality or HF event, and LV volume reduction at 12 months. Tertiary endpoints include changes in quality of life, NYHA functional class, 6 min walk test, natriuretic peptides, and safety outcomes. The results of our prospective, randomized, multicentre clinical trial will provide important information on the role of cardiac resynchronization therapy with defibrillator (CRT-D) upgrade in patients with symptomatic HF, reduced LVEF, and wide-paced QRS with intermittent or permanent RV pacing. NCT02270840. © The Author 2016. Published by Oxford University Press on behalf of the European Society of Cardiology.

  12. Upgrading during difficult times

    International Nuclear Information System (INIS)

    Tiefenbach, K.

    1993-01-01

    The Co-op Upgrader is part of an integrated refinery and upgrader complex in Regina, Saskatchewan. The upgrader processes 50,000 bbl/d heavy sour crude oil, mostly Fosterton and dilute Lloydminster crude, via hydrogen addition and carbon rejection, desulfurization, demetallization, and denitrification to yield a synthetic crude blend. The synthetic crude is refined to produce gasoline and diesel fuel. Byproducts from the integrated operation include 100,000 tonnes/y of petroleum coke, 65,000 tonnes/y of sulfur, propane, butane, fuel oil, and metals (Ni and V) in the form of spent catalysts. Recent operational and economic challenges faced by the upgrader are reviewed. Technical challenges include operating the upgrader's high-temperature high-pressure heavy oil hydrotreating unit and distillate hydrogenation unit, removal and replacement of the desulfurization catalyst, waste management, and producing coke of sufficient quality. Economic challenges include the shrinking differential between light and heavy oil, higher prices for natural gas (the main raw material for hydrogen production for upgrading), seasonal changes in product specifications, and lower prices for sulfur and metal byproducts. The upgrader is also affected by interest rates since borrowing costs are the single largest expenditure after crude oil purchases. 4 figs

  13. Using the Human Systems Simulation Laboratory at Idaho National Laboratory for Safety Focused Research

    Energy Technology Data Exchange (ETDEWEB)

    Joe, Jeffrey .C; Boring, Ronald L.

    2016-07-01

    Under the United States (U.S.) Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) program, researchers at Idaho National Laboratory (INL) have been using the Human Systems Simulation Laboratory (HSSL) to conduct critical safety focused Human Factors research and development (R&D) for the nuclear industry. The LWRS program has the overall objective to develop the scientific basis to extend existing nuclear power plant (NPP) operating life beyond the current 60-year licensing period and to ensure their long-term reliability, productivity, safety, and security. One focus area for LWRS is the NPP main control room (MCR), because many of the instrumentation and control (I&C) system technologies installed in the MCR, while highly reliable and safe, are now difficult to replace and are therefore limiting the operating life of the NPP. This paper describes how INL researchers use the HSSL to conduct Human Factors R&D on modernizing or upgrading these I&C systems in a step-wise manner, and how the HSSL has addressed a significant gap in how to upgrade systems and technologies that are built to last, and therefore require careful integration of analog and new advanced digital technologies.

  14. Safety Analysis in Design and Assessment of the Physical Protection of the OKG NPP

    Energy Technology Data Exchange (ETDEWEB)

    Lindahl, P., E-mail: par.lindahl@okg.eon.se [OKG Aktiebolag, Oskarshamn (Sweden)

    2014-10-15

    OKG AB operates a three unit nuclear power plant in the southern parts of Sweden. As a result of recent development of the legislation regarding physical protection of nuclear facilities, OKG has upgraded the protection against antagonistic actions. The new legislation includes requirements both on specific protective measures and on the performance of the physical protection as a whole. In short, the performance related requirements state that sufficient measures shall be implemented to protect against antagonistic actions, as defined by the regulator in the “Design Basis Threat” (DBT). Historically, physical protection and nuclear safety has been managed much as separate issues with different, sometimes contradicting, objectives. Now, insights from the work with the security upgrade have emphasized that physical protection needs to be regarded as an important part of the Defence-In-Depth (DiD) against nuclear accidents. Specifically, OKG has developed new DBT-based analysis methods, which may be characterized as probabilistically informed deterministic analysis, conformed to a format similar to the one used for conventional internal events analysis. The result is a powerful tool for design and assessment of the performance of the protection against antagonistic actions, using a nuclear safety perspective. (author)

  15. Technical evaluation report of the Fort St. Vrain final draft upgraded technical specifications

    International Nuclear Information System (INIS)

    Kimura, C.Y.

    1989-01-01

    This report is a technical evaluation of the final draft of the Fort St. Vrain (FSV) Upgraded Technical Specifications (UT/S) as issued by Public Service of Colorado (PSC) on May 27, 1988 with subsequent supplemental updates issued on June 15, 1988 and August 5, 1988. It has been compared for consistency, and safety conservatism with the Fort St. Vrain (FSV) Updated Final Safety Analysis Report (FSAR), the FSV Safety Evaluation Report (SER), the Facility Operating License, DPR-34, and all amendments to the Facility Operating License issued as of June 1, 1988, and Appendix A to the Operating License DPR-34, Technical Specifications. Because of the age of the plant, no supplements to the Fort St. Vrain SER have been issued since the original SER was not issued as a WASH or a NUREG report. This made it necessary to review all amendments to the Facility Operating License since they would contain the safety evaluations done to support changes to the Facility Operating License. The upgraded Fort St. Vrain Technical Specifications were also broadly compared with the latest Westinghouse Standard Technical Specifications (WSTS) to assure that what was proposed for Fort St. Vrain was consistent with the latest NRC staff practices for standard technical specifications

  16. NSLS control system upgrade

    International Nuclear Information System (INIS)

    Smith, J.D.; Ramamoorthy, Susila; Tang, Y.N.

    1994-01-01

    The NSLS consists of two storage rings, a booster and a linac. A major upgrade of the control system (installed in 1978) was undertaken and has been completed. The computer architecture is being changed from a three level star-network to a two level distributed system. The microprocessor subsystem, host computer and workstations, communication link and the main software components are being upgraded or replaced. Since the NSLS rings operate twenty four hours a day a year with minimum maintenance time, the key requirement during the upgrade phase is a non-disruptive transition with minimum downtime. Concurrent with the upgrade, some immediate improvements were required. This paper describes the various components of the upgraded system and outlines the future plans. ((orig.))

  17. NSLS control system upgrade

    International Nuclear Information System (INIS)

    Smith, J.D.; Ramamoorthy, S.; Tang, Yong N.

    1995-01-01

    The NSLS consists of two storage rings, a booster and a linac. A major upgrade of the control system (installed in 1978) was undertaken and has been completed. The computer architecture is being changed from a three level star-network to a two level distributed system. The microprocessor subsystem, host computer and workstations, communication link and the main software components are being upgraded or replaced. Since the NSLS rings operate twenty four hours a day a year with minimum maintenance time, the key requirement during the upgrade phase is a non-disruptive transition with minimum downtime. Concurrent with the upgrade, some immediate improvements were required. This paper describes the various components of the upgraded system and outlines the future plans

  18. Bi-Provincial Upgrader

    International Nuclear Information System (INIS)

    1997-01-01

    Husky Oil's Bi-Provincial Upgrader is located in the rural municipality of Wilton, east of Lloydminster, Saskatchewan. It is jointly owned by Husky Oil and the Saskatchewan government. The upgrader is designed to produce 7.300 m 3 /cd of synthetic crude from 8.440 m 3 /cd of diluted Lloyd blend and distillates from the existing Lloydminster refinery. The designed coke and sulphur production is 415 t/day and 240 t/day respectively. Chronology of the Bi-Provincial Upgrader project was presented, along with details of the heavy oil feedstock properties for Cold Lake, Lloydminster and Lloyd Blend. Upgrader production history since start-up, synthetic crude oil customers, and the evolution of prices for Edmonton light vs. Husky LLB at Hardisty were also reviewed. 3 tabs., 9 figs

  19. Concept for creating program-technical complex of safety monitoring with system of safety parameters presentation functions on the basis of routine WWER-1000 systems

    International Nuclear Information System (INIS)

    Dunaev, V.G.; Tarasov, M. V.; Povarov, P.V.

    2005-01-01

    Prerequisites of creating the software-hardware complex for reactor safety monitoring on the Volgodonsk NPP are analyzed and generalized. The concept of this complex is based on functions of the safety parameters presentation system. It will serve as an interface between operator and technological process and give to operator a possibility to estimate quickly the state of the safety of the nuclear power unit. The complex will be created on the basis of routine reactor monitoring and control systems intended for the WWER-1000 reactor. In addition to existing soft- and hard-wares for reactor monitoring and for analysis of technological archive, it is proposed to create and connect in parallel the new software-hardware complex which ensures calculation and presentation of generalized factors of reactor safety [ru

  20. Upgrade of the Annular Core Pulse Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Reuscher, J A [Sandia Laboratories, Albuquerque, NM (United States)

    1976-07-01

    The Annular Core Pulse Reactor (ACPR) is a TRIGA type reactor which has been in operation at Sandia Laboratories since 1967. The reactor is utilized in a wide variety of experimental programs which include radiation effects, neutron radiography, activation analysis, and fast reactor safety. During the past two years, the ACPR has become an important experimental facility for the United States Fast Reactor Safety Research Program and questions of interest to the safety of the LMFBR are being addressed. In order to enhance the capabilities of the ACPR for reactor safety experiments, a project to improve the performance of the reactor was initiated. It is anticipated that the pulse fluence can be increased by a factor of 2.0 to 2.5 by utilizing a two-region core concept with high heat capacity fuel elements around the central irradiation cavity. In addition, the steady-state power of the reactor will be increased by about a factor of two. Preliminary studies have identified several potential approaches to the ACPR performance improvement. The most promising approach appears to be the two-region core concept. The inner region, surrounding the irradiation cavity, would consist of a high-heat capacity fuel capable of absorbing the fission energy associated with a large nuclear pulse. The number of fissions occurring near the cavity would be greatly increased which, in turn, would significantly increase the fluence in the cavity. The outer region would consist of a U-ZrH fuel to provide an overall negative temperature coefficient for the two region core. Two candidate high heat capacity fuels [(BeO-UO{sub 2} and UC-ZrC) - graphite] are under consideration. Since this reactor upgrade represents a modification to an existing facility, it can be achieved in a relatively short time. It is anticipated that most of the existing reactor structure can be used for the upgrade. The present core occupies about one-half of the location in the grid plate. The high-heat capacity fuel

  1. Incidence, predictors, and procedural results of upgrade to resynchronization therapy: the RAFT upgrade substudy.

    Science.gov (United States)

    Essebag, Vidal; Joza, Jacqueline; Birnie, David H; Sapp, John L; Sterns, Laurence D; Philippon, Francois; Yee, Raymond; Crystal, Eugene; Kus, Teresa; Rinne, Claus; Healey, Jeffrey S; Sami, Magdi; Thibault, Bernard; Exner, Derek V; Coutu, Benoit; Simpson, Chris S; Wulffhart, Zaev; Yetisir, Elizabeth; Wells, George; Tang, Anthony S L

    2015-02-01

    The resynchronization-defibrillation for ambulatory heart failure trial (RAFT) study demonstrated that adding cardiac resynchronization therapy (CRT) in selected patients requiring de novo implantable cardiac defibrillators (ICD) reduced mortality as compared with ICD therapy alone, despite an increase in procedure-related adverse events. Data are lacking regarding the management of patients with ICD therapy who develop an indication for CRT upgrade. Participating RAFT centers provided data regarding de novo CRT-D (CRT with ICD) implant, upgrade to CRT-D during RAFT (study upgrade), and upgrade within 6 months after presentation of study results (substudy). Substudy centers enrolled 1346 (74.9%) patients in RAFT, including 644 de novo, 80 study upgrade, and 60 substudy CRT attempts. The success rate (initial plus repeat attempts) was 95.2% for de novo versus 96.3% for study upgrade and 90.0% for substudy CRT attempts (P=0.402). Acute complications occurred among 26.2% of de novo versus 18.8% of study upgrade and 3.4% of substudy CRT implantation attempts (PRAFT study and other trials. © 2014 American Heart Association, Inc.

  2. Seismic safety of nuclear power plants

    International Nuclear Information System (INIS)

    Guerpinar, A.; Godoy, A.

    2001-01-01

    This paper summarizes the work performed by the International Atomic Energy Agency in the areas of safety reviews and applied research in support of programmes for the assessment and enhancement of seismic safety in Eastern Europe and in particular WWER type nuclear power plants during the past seven years. Three major topics are discussed; engineering safety review services in relation to external events, technical guidelines for the assessment and upgrading of WWER type nuclear power plants, and the Coordinated Research Programme on 'Benchmark study for the seismic analysis and testing of WWER type nuclear power plants'. These topics are summarized in a way to provide an overview of the past and present safety situation in selected WWER type plants which are all located in Eastern European countries. Main conclusion of the paper is that although there is now a thorough understanding of the seismic safety issues in these operating nuclear power plants, the implementation of seismic upgrades to structures, systems and components are lagging behind, particularly for those cases in which the re-evaluation indicated the necessity to strengthen the safety related structures or install new safety systems. (author)

  3. Regulatory standpoints on the design-basis capability of safety-related motor-operated valves(MOVs) and power-operated gate valves(POGVs)

    International Nuclear Information System (INIS)

    Kim, W. T.; Kum, O. H.

    1999-01-01

    The weakness in the design-basis capability of Motor-Operated Valves(MOVs) and the susceptibility to Pressure Locking and Thermal Binding phenomena of Power-Operated Gate Valves(POGVs) have been major concerns to be resolved in the nuclear society in and abroad since Three Mile Island accident occurred in the USA in 1979. Through detailed analysis of operating experience and regulatory activities, some MOVs and POGVs have been found to be unreliable in performing their safety functions when they are required to do so under certain conditions, especially under design-basis accident conditions. Further, it is well understood that these safety problems may not be identified by the typical valve in-service testing(IST). USNRC has published three Generic Letters, GL 89-10, GL 95-07, and GL 96-05, requiring nuclear plant licensees to take appropriate actions to resolve the problems mentioned above. Korean nuclear regulatory body has made public an administration measure called 'Regulatory recommendation to verify safety functions of the safety-related MOVs and POGVs' on June 13, 1997, and in this administration measure Korean utility is asked to submit written documents to show how it assure design-basis capability of these valves. The following are among the major concerns being considered from a regulation standpoint. Program scope and implementation priority, dynamic tests under differential pressure conditions, accuracy of diagnostic equipment, torque switch setting and torque bypass percentage, weak link analysis, motor actuator sizing, corrective actions taken to resolve pressure locking and thermal binding susceptibility, and a periodic verification program for the valves once design-basis capability has been verified

  4. Evaluation of safety assessment methodologies in Rocky Flats Risk Assessment Guide (1985) and Building 707 Final Safety Analysis Report (1987)

    International Nuclear Information System (INIS)

    Walsh, B.; Fisher, C.; Zigler, G.; Clark, R.A.

    1990-01-01

    FSARs. Rockwell International, as operating contractor at the Rocky Flats plant, conducted a safety analysis program during the 1980s. That effort resulted in Final Safety Analysis Reports (FSARs) for several buildings, one of them being the Building 707 Final Safety Analysis Report, June 87 (707FSAR) and a Plant Safety Analysis Report. Rocky Flats Risk Assessment Guide, March 1985 (RFRAG85) documents the methodologies that were used for those FSARs. Resources available for preparation of those Rocky Flats FSARs were very limited. After addressing the more pressing safety issues, some of which are described below, the present contractor (EG ampersand G) intends to conduct a program of upgrading the FSARs. This report presents the results of a review of the methodologies described in RFRAG85 and 707FSAR and contains suggestions that might be incorporated into the methodology for the FSAR upgrade effort

  5. Control processes and machine protection on ASDEX Upgrade

    International Nuclear Information System (INIS)

    Raupp, G.; Treutterer, W.; Mertens, V.; Neu, G.; Sips, A.; Zasche, D.; Zehetbauer, Th.

    2007-01-01

    Safe operation of ASDEX Upgrade is guaranteed by a conventional hierarchy of simple and robust hard-wired systems for personnel and machine protection featuring standardized switch-off procedures. Machine protection and handling of off-normal events is further enhanced and peak and lifetime stress minimized through the plasma control system. Based on a real-time process model supporting safety critical applications with data quality tagging, process self-monitoring, watchdog monitoring and alarm propagation, processes detect complex and critical failures and reliably perform case-sensitive counter measures. Intelligent real-time failure handling is done with hardware or software redundancy and performance degradation, or modification of reference values to continue or terminate discharges with reduced machine stress. Examples implemented so far on ASDEX Upgrade are given, such as recovery from measurement failures, switch-over of redundant actuators, handling of actuator limitations, detection of plasma instabilities, plasma state dependent soft landing, or handling of failed switch-off procedures through breakers disconnecting the machine from grid

  6. Safety Assessment Methodologies and Their Application in Development of Near Surface Waste Disposal Facilities--ASAM Project

    International Nuclear Information System (INIS)

    Batandjieva, B.; Metcalf, P.

    2003-01-01

    Safety of near surface disposal facilities is a primary focus and objective of stakeholders involved in radioactive waste management of low and intermediate level waste and safety assessment is an important tool contributing to the evaluation and demonstration of the overall safety of these facilities. It plays significant role in different stages of development of these facilities (site characterization, design, operation, closure) and especially for those facilities for which safety assessment has not been performed or safety has not been demonstrated yet and the future has not been decided. Safety assessments also create the basis for the safety arguments presented to nuclear regulators, public and other interested parties in respect of the safety of existing facilities, the measures to upgrade existing facilities and development of new facilities. The International Atomic Energy Agency (IAEA) has initiated a number of research coordinated projects in the field of development and improvement of approaches to safety assessment and methodologies for safety assessment of near surface disposal facilities, such as NSARS (Near Surface Radioactive Waste Disposal Safety Assessment Reliability Study) and ISAM (Improvement of Safety Assessment Methodologies for Near Surface Disposal Facilities) projects. These projects were very successful and showed that there is a need to promote the consistent application of the safety assessment methodologies and to explore approaches to regulatory review of safety assessments and safety cases in order to make safety related decisions. These objectives have been the basis of the IAEA follow up coordinated research project--ASAM (Application of Safety Assessment Methodologies for Near Surface Disposal Facilities), which will commence in November 2002 and continue for a period of three years

  7. New operational spaces for the electron cyclotron resonance heating at ASDEX upgrade

    International Nuclear Information System (INIS)

    Hoehnle, Hendrik Sebastian

    2012-01-01

    In this thesis, new electron cyclotron resonance heating (ECRH) scenarios were developed for an extension of the operational space at the tokamak ASDEX Upgrade in view of ITER compatibility. In the last years, the first wall material at ASDEX Upgrade was changed from graphite to tungsten, and the ECRH is needed to control the tungsten concentration in the plasma core. But, in ITER-like plasma discharges at ASDEX Upgrade, the usage of the ECRH in the typically used second harmonic extraordinary polarised mode (X2 mode) is limited. In these ITER-scenarios a small safety factor should be achieved, which implements an increase of the plasma current at ASDEX Upgrade. A higher plasma current and a high confinement lead to a raised density and for the ITER scenario to an electron density above the cutoff of the X2 mode at ASDEX Upgrade. Therefore, the X2 mode is reflected at the cutoff layer and cannot be used for central heating and the control of the tungsten concentration. One possibility to overcome this problem is to apply the third harmonic mode at reduced magnetic field. Here the cutoff is increased by 33% due to the dependence on the magnetic field. However, at the reachable plasma parameters at the reduced field the absorption of the X3 mode is incomplete (60-70 %) and the shine-trough power can destroy microwave sensitive components in ASDEX Upgrade. To solve this problem the magnetic field has to be optimized. A slightly increased magnetic field from 1.7 T to 1.8 T moves the second harmonic resonance in the region of confined plasma with high temperatures and density, so that this resonance can act as beam dump. The deposition in the plasma core is still central enough for the tungsten control ability of the ECRH. The benefit of the beam dump was verified in experiments with two different magnetic fields (1.7 T and 1.8 T). In case of the higher magnetic field, the stray radiation was reduced; simultaneously the electron temperature was increased. In addition

  8. The D0 Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Abachi, S.; D0 Collaboration

    1995-07-01

    In this paper we describe the approved DO Upgrade detector, and its physics capabilities. The DO Upgrade is under construction and will run during the next Fermilab collider running period in early 1999 (Run II). The upgrade is designed to work at the higher luminosities and shorter bunch spacings expected during this run. The major elements of t he upgrade are: a new tracking system with a silicon tracker, scintillating fiber tracker, a 2T solenoid, and a central preshower detector; new calorimeter electronics; new muon trigger and tracking detectors with new muon system electronics; a forward preshower detector; new trigger electronics and DAQ improvements to handle the higher rates.

  9. Notes on safety modernization of nuclear power plants in Sweden

    International Nuclear Information System (INIS)

    Hammar, L.

    1997-01-01

    Investment programmes are pursued for all reactor generations in the order of 70M USD per year and unit, despite the political decision to phase out nuclear power. 15-20% of this may be safety-related. Major redesign and replacements of piping and joints in the primary system of Ringhals-1 (1:st generation BWR, external pump loops) are under-way aimed at enhancing the barrier reliability at par with with the new reactors with internal circulation pumps. Major upgrading in process control are typically on the agenda, e.g. modern digital protection and control systems as installed in Ringhals 1 and 2 in 1995. Comprehensive modernization of the control rooms are planned for all Swedish reactors, commencing in 1997 with the Forsmark reactors. The needs for modernization in regard of safety are expected to be further clarified in the design basis reviews due for completion in 1998

  10. Status of TMX upgrade diagnostics construction

    International Nuclear Information System (INIS)

    Hornady, R.S.; Davis, J.C.; Simonen, T.C.

    1981-01-01

    This report describes the status of the initial TMX Upgrade diagnostics and the state of development of additional diagnostics being prepared for later TMX Upgrade experiments. The initial diagnostic instrument set has been described in the TMX Upgrade Proposal. This set is required to get TMX Upgrade operational and to evaluate its initial performance. Additional diagnostic instruments are needed to then carry out the more detailed experiments outlined by the TMX Upgrade program milestones. The relation of these new measurements to the physics program is described in The TMX Upgrade Program Plan

  11. THE ROLE OF NAVIGATIONAL AIDS IN FLIGHT SAFETY MANAGEMENT WITHIN ICAO GLOBAL AIR NAVIGATION PLAN

    Directory of Open Access Journals (Sweden)

    Vadim V. Vurobyov

    2017-01-01

    Full Text Available The development of the global civil aviation is provided on the basis of the ICAO Communication and Surveillance/Air Traffic Management Concept, which has determined the basic strategy for further commercial flight management effectiveness improvement. On the basis of this concept a Global Air Navigation Plan has been developed by ICAO recently. The core strategies of CNS/ATM concept were specified and combined into so-called blocks. Thus the term Global Aviation System block upgrade has been introduced. At the same time, GANP states that the introduction of new procedures and flight management systems will inevitably affect flight safety. Accordingly, there is a task of flight safety management level maintaining, or even increasing within the Global Air Navigation Plan implementation. Various air navigational aids play a significant role in the process as they are directly associated with the new systems and structures introduction.This breeds the new global challenge of flight safety management level change assessment during the introduction of new procedures and systems connected with the use of both navigational aids and instruments. Some aspects of this problem solution are covered in the article.

  12. Slum Upgrading and Health Equity.

    Science.gov (United States)

    Corburn, Jason; Sverdlik, Alice

    2017-03-24

    Informal settlement upgrading is widely recognized for enhancing shelter and promoting economic development, yet its potential to improve health equity is usually overlooked. Almost one in seven people on the planet are expected to reside in urban informal settlements, or slums, by 2030. Slum upgrading is the process of delivering place-based environmental and social improvements to the urban poor, including land tenure, housing, infrastructure, employment, health services and political and social inclusion. The processes and products of slum upgrading can address multiple environmental determinants of health. This paper reviewed urban slum upgrading evaluations from cities across Asia, Africa and Latin America and found that few captured the multiple health benefits of upgrading. With the Sustainable Development Goals (SDGs) focused on improving well-being for billions of city-dwellers, slum upgrading should be viewed as a key strategy to promote health, equitable development and reduce climate change vulnerabilities. We conclude with suggestions for how slum upgrading might more explicitly capture its health benefits, such as through the use of health impact assessment (HIA) and adopting an urban health in all policies (HiAP) framework. Urban slum upgrading must be more explicitly designed, implemented and evaluated to capture its multiple global environmental health benefits.

  13. Health risk from radioactive and chemical environmental contamination: common basis for assessment and safety decision making

    International Nuclear Information System (INIS)

    Demin, V.

    2004-01-01

    To meet the growing practical need in risk analysis in Russia health risk assessment tools and regulations have been developed in the frame of few federal research programs. RRC Kurchatov Institute is involved in R and D on risk analysis activity in these programs. One of the objectives of this development is to produce a common, unified basis of health risk analysis for different sources of risk. Current specific and different approaches in risk assessment and establishing safety standards developed for chemicals and ionising radiation are analysed. Some recommendations are given to produce the common approach. A specific risk index R has been proposed for safety decision-making (establishing safety standards and other levels of protective actions, comparison of various sources of risk, etc.). The index R is defined as the partial mathematical expectation of lost years of healthy life (LLE) due to exposure during a year to a risk source considered. The more concrete determinations of this index for different risk sources derived from the common definition of R are given. Generic safety standards (GSS) for the public and occupational workers have been suggested in terms of this index. Secondary specific safety standards have been derived from GSS for ionizing radiation and a number of other risk sources including environmental chemical pollutants. Other general and derived levels for decision-making have also been proposed including the e-minimum level. Their possible dependence on the national or regional health-demographic data is shortly considered. Recommendations are given on methods and criteria for comparison of various sources of risk. Some examples of risk comparison are demonstrated in the frame of different comparison tasks. The paper has been prepared on the basis of the research work supported by International Science and Technology Centre, Moscow (project no. 2558). (author)

  14. FLIGHT SAFETY CONTROL OF THE BASIS OF UNCERTAIN RISK EVALUATION WITH NON-ROUTINE FLIGHT CONDITIONS INVOLVED

    Directory of Open Access Journals (Sweden)

    2016-01-01

    Full Text Available The article deals with methods of forecasting the level of aviation safety operation of aircraft systems on the basis of methods of evaluation the risks of negative situations as a consequence of a functional loss of initial properties of the system with critical violations of standard modes of the aircraft. Mathematical Models of Risks as a Danger Measure of Discrete Random Events in Aviation Systems are presented. Technological Schemes and Structure of Risk Control Proce- dures without the Probability are illustrated as Methods of Risk Management System in Civil Aviation. The assessment of the level of safety and quality and management of aircraft, made not only from the standpoint of reliability (quality and consumer properties, but also from the position of ICAO on the basis of a risk-based approach. According to ICAO, the security assessment is performed by comparing the calculated risk with an acceptable level. The approach justifies the use of qualitative evaluation techniques safety in the forms of proactive forecasted and predictive risk management adverse impacts to aviation operations of various kinds, including the space sector and nuclear energy. However, for the events such as accidents and disasters, accidents with the aircraft, fighters in a training flight, during the preparation of the pilots on the training aircraft, etc. there is no required statistics. Density of probability distribution (p. d. f. of these events are only hypothetical, unknown with "hard tails" that completely eliminates the application of methods of confidence intervals in the traditional approaches to the assessment of safety in the form of the probability analysis.

  15. Acceptance criteria for the physical protection upgrade rule requirements for fixed sites. Information guide

    International Nuclear Information System (INIS)

    Dwyer, P.

    1980-09-01

    This document has been developed as a tool to assist in providing consistent evaluation of upgraded physical security plans submitted in response to the Physical Protection Upgrade Rule, effective March 25, 1980. It presents a means for assuring licensee compliance with every regulatory requirement of particular significance to the protection of the public health and safety. Acceptance criteria are included to determine the extent to which each licensee meets the regulatory requirements. The format parallels Regulatory Guide 5.52, Standard Format and Content of a Licensee Physical Protection Plan for Strategic Special Nuclear Material at Fixed Sites

  16. Core thermohydraulic design with LEU fuels for upgraded research reactor, JRR-3

    Energy Technology Data Exchange (ETDEWEB)

    Sudo, Y; Ando, H; Ikawa, H; Ohnishi, N [Department of Research Reactor Operation, Japan Atomic Energy Research Institute (JAERI), 319-11 Tokai-Mura, Ibaraki-Ken (Japan)

    1985-07-01

    This paper presents the outline of core thermohydraulic design and analysis of the research reactor, JRR-3, which is to be upgraded to a 20 MWt pool-type, light water-cooled reactor with 20% LEU plate-type fuels. The major feature of core thermohydraulics of the upgraded JRR-3 is that core flow is a downflow at the condition of normal operation, with which fuel plates are exposed to a severer condition than with an upflow in case of operational transients and accidents. The core thermo-hydraulic design was, therefore, done for the condition of normal operation so that fuel plates may have enough safety margin both against the onset of nucleate boiling not to allow the nucleate boiling anywhere in the core and against the initiation of DNB, and the safety margin for these were evaluated. The core velocity thus designed is at the optimum condition where fuel plates have the maximum margin against the onset of nucleate boiling. The core thermohydraulic characteristics were also clarified for the natural circulation cooling mode. (author)

  17. Hardware and software maintenance strategies for upgrading vintage computers

    International Nuclear Information System (INIS)

    Wang, B.C.; Buijs, W.J.; Banting, R.D.

    1992-01-01

    The paper focuses on the maintenance of the computer hardware and software for digital control computers (DCC). Specific design and problems related to various maintenance strategies are reviewed. A foundation was required for a reliable computer maintenance and upgrading program to provide operation of the DCC with high availability and reliability for 40 years. This involved a carefully planned and executed maintenance and upgrading program, involving complementary hardware and software strategies. The computer system was designed on a modular basis, with large sections easily replaceable, to facilitate maintenance and improve availability of the system. Advances in computer hardware have made it possible to replace DCC peripheral devices with reliable, inexpensive, and widely available components from PC-based systems (PC = personal computer). By providing a high speed link from the DCC to a PC, it is now possible to use many commercial software packages to process data from the plant. 1 fig

  18. Regulatory requirements for replacement of analog systems with digital upgrades

    International Nuclear Information System (INIS)

    Loeser, P.J.

    1993-01-01

    This paper reviews briefly the regulatory guidelines which must be met in order to replace analog systems in nuclear power plants with digital systems. There is a move to do such replacements for a number of reasons: analog systems are aging, and showing considerable drift; few vendors manufacture analog systems today; support and parts are hard to get; digital systems provide flexibility. There is a safety concern however about undesirable and unpredictable effects to digital safety equipment due to plant transients, accidents, post-accident condition, and EMI/RF environmental interferences. License holders must comply with the requirements of 10 C.F.R. 50.59, which deals with safety concerns with respect to any changes to operating plants which may have an impact on the safety of the plant. NRC staff is taking the position that all digital upgrades will require an evaluation under this regulation

  19. System architecture of Detector Control and safety for the ATLAS Inner Detector Upgrade

    International Nuclear Information System (INIS)

    Ferrere, D.; Kersten, S.

    2011-01-01

    In the current ATLAS Upgrade plan a new Inner Detector (ID) based upon silicon sensor technology is being considered. The operational monitoring and control of the ID will be very demanding. The Detector Control System (DCS) is a common tool that is essential for the operational safety of a system. Even at this early stage the DCS system architecture has to be defined such that it is well integrated and optimized for its later implementation and use. For example the DCS diagnostics for the front-end (FE) chips is a serious option being considered that needs an early requirement and specification definition. In addition one of the main constraints is the service reuse between the service patch panels of the ATLAS ID and the counting room that limits the number of electrical lines to be reused. Conceptual differences in terms of readout architecture and layout have been identified between the strip and the pixel detector that lead to two distinct architectures. Nevertheless, the limitation of available electrical lines going to the counting room as well as the low material budget requirements inside the ID volume are two major constraints that lead the ID to consider an on-detector radiation hard integrated circuitry for the slow control. At this stage of the project, the definitions of the logical actions and protocol for the ADCs of such a chip are still being specified. In addition the experience gained from the current ID will be essential for the guidance of tuning the future DCS architecture in the coming years.

  20. Safety of RBMK reactors: Major results and prospects

    International Nuclear Information System (INIS)

    Sidorenko, V.A.

    1996-01-01

    The paper considers the following issues: basic reasons for the advent of NPPs with RBMK reactors; the logic of identifying top-priority measures immediately after the accident; top-priority measures for improving the safety and reliability of NPPs with RBMK reactors; upgrading NPPs with RBMK reactors in compliance with the Norms; programmes for retrofitting and upgrading of NPPs of the ''Rosnergoatom'' Concern and progress with their implementation as of April 1996; the safety of RBMK plants and the programmes of its enhancement with regard to modern requirements in the light of national and international assessment; objective indicators of safety, reliability, and economic efficiency of NPPs with RBMK reactors; economics: rationale for continuing plants operation till the end of their design lifetime. 8 refs, 3 figs

  1. Limiter and divertor systems - conceptual and mechanical design for Aditya Tokamak upgrade

    International Nuclear Information System (INIS)

    Patel, Kaushal; Rathod, Kulav; Jadeja, Kumarpalsinh A.

    2015-01-01

    Existing Aditya tokamak with limiter configuration is being upgraded into a machine to have both the limiter and divertor configurations. Necessary modifications have been carried out to accommodate divertor coils by replacing the old vacuum vessel with a new circular section vacuum vessel. The upgraded Aditya tokamak will have different set of limiters and divertors, such as Safety limiter, Toroidal Inner limiter, outer limiter of smaller toroidal extent, Upper and lower divertor plates. The limiter and divertor locations inside the Aditya tokamak upgrade are decided based on the numerical simulation of the plasma equilibrium profiles. Initially graphite will be used as plasma facing material (PFM) in all the limiter and divertor plates. The dimensions of the limiter and divertor tiles are decided based on their installation inside the vacuum vessel as well as on the total plasma heat loads (∼ 1 MW) falling on them. Depending upon the heat loads; the thickness of graphite tiles for limiter and divertor plates is estimated. Shaped graphite tiles will be fixed on specially designed support structures made out of SS-304L inside the torus shaped vacuum vessel. In this paper mechanical structural design of limiter and divertor of Aditya Upgrade Tokamak is presented. (author)

  2. Digital upgrade of radiation-monitoring-system subcomponents

    International Nuclear Information System (INIS)

    Bohrisch, R.L

    1993-01-01

    This paper describes the experience of Southern California Edison (SCE) in upgrading an obsolete, analog, printed circuit board contain in most of the process and effluent radiation detectors at the San Onofre Nuclear Generating Station. The printed circuit board, which functions to produce a linear voltage and current that is proportional to the log of the radiation level, was reengineered by SCE with microprocessor-based digital technology and subjected to qualification testing, including seismic and environmental, for use in class I safety-related applications. The results, benefits, and disadvantages to this approach are discussed in this paper

  3. Safety culture in nuclear installations. Management of safety and safety culture in Indian NPPs

    International Nuclear Information System (INIS)

    Rawal, S.C.

    2002-01-01

    Nuclear Power Corporation Of India Ltd. (NPCIL) is a company owned by Government of India and is responsible for Design, Construction, Commissioning, Operation and Decommissioning of Nuclear Power plants in India. Presently, a total of 13 Nuclear power Stations are in operation with an installed capacity of 2620 MWe and 2 VVR type PWR Units of 1000 MWe capacity each, 2 PHWR type units of 500 MWe capacity each and 4 PHWR type 220 MWe capacity each are under construction. NPPs generation capacity has been increased from 70% to 85% in the span Of last 7 years with high level of safety standards. This could be achieved through Management commitment towards building a strong Safety Culture. Safety culture is that assembly of characteristics and attitudes in organisation and individuals which establishes that as an overriding priority nuclear plant safety issues receives the attention warranted by their significance. This definition of safety culture brings out two major components in its manifestation. The framework within which individuals within the organisation works.The attitude and response of individual towards the safety issues over productivity and economics in the organisational work practices. The two attributes of safety culture are built in and upgraded in each individuals through special training at the time of entry in the organisation and later through in built procedures in the work practices, motivation and encouragement for free participation of each individuals. Individuals are encouraged to participate in Quality circle teams at the sectional level and review of safety proposal originated by individuals in Station operation Review Committee at Station level, in addition to this to continuously enhance the safety culture, refresher training courses are being organised at regular intervals. The safety related proposals are categorised in to two namely: Proposals from Operating Plants, and Proposals from projects and Design. The concept of safety

  4. Repairing and Upgrading Your PC

    CERN Document Server

    Thompson, Robert

    2009-01-01

    Repairing and Upgrading Your PC delivers start-to-finish instructions, simple enough for even the most inexperienced PC owner, for troubleshooting, repairing, and upgrading your computer. Written by hardware experts Robert Bruce Thompson and Barbara Fritchman Thompson, this book covers it all: how to troubleshoot a troublesome PC, how to identify which components make sense for an upgrade, and how to tear it all down and put it back together. This book shows how to repair and upgrade all of your PC's essential components.

  5. Abnormal condition and events analysis for instrumentation and control systems. Volume 1: Methodology for nuclear power plant digital upgrades. Final report

    International Nuclear Information System (INIS)

    McKemy, S.; Marcelli, M.; Boehmer, N.; Crandall, D.

    1996-01-01

    The ACES project was initiated to identify a cost-effective methodology for addressing abnormal conditions and events (ACES) in digital upgrades to nuclear power plant systems, as introduced by IEEE Standard 7-4.3.2-1993. Several methodologies and techniques currently in use in the defense, aerospace, and other communities for the assurance of digital safety systems were surveyed, and although several were shown to possess desirable qualities, non sufficiently met the needs of the nuclear power industry. This report describes a tailorable methodology for performing ACES analysis that is based on the more desirable aspects of the reviewed methodologies and techniques. The methodology is applicable to both safety- and non-safety-grade systems, addresses hardware, software, and system-level concerns, and can be applied in either a lifecycle or post-design timeframe. Employing this methodology for safety systems should facilitate the digital upgrade licensing process

  6. Jefferson Lab IEC 61508/61511 Safety PLC Based Safety System

    International Nuclear Information System (INIS)

    Mahoney, Kelly; Robertson, Henry

    2009-01-01

    This paper describes the design of the new 12 GeV Upgrade Personnel Safety System (PSS) at the Thomas Jefferson National Accelerator Facility (TJNAF). The new PSS design is based on the implementation of systems designed to meet international standards IEC61508 and IEC 61511 for programmable safety systems. In order to meet the IEC standards, TJNAF engineers evaluated several SIL 3 Safety PLCs before deciding on an optimal architecture. In addition to hardware considerations, software quality standards and practices must also be considered. Finally, we will discuss R and D that may lead to both high safety reliability and high machine availability that may be applicable to future accelerators such as the ILC.

  7. NPP Mochovce nuclear safety enhancement program

    International Nuclear Information System (INIS)

    Cech, J.; Baumester, P.

    1997-01-01

    Nuclear power plant Mochovce is currently under construction and an extensive nuclear safety enhancement programme is under way. The upgrading and modifications are based on IAEA documents and on those of the Nuclear Regulatory Authority of the Slovak Republic. Based on a contract concluded with Riskaudit from the CEC, safety examinations of the Mochovce design were performed. An extensive list of technical specifications of safety measures is given. (M.D.)

  8. 242-A evaporator safety analysis report

    International Nuclear Information System (INIS)

    CAMPBELL, T.A.

    1999-01-01

    This report provides a revised safety analysis for the upgraded 242-A Evaporator (the Evaporator). This safety analysis report (SAR) supports the operation of the Evaporator following life extension upgrades and other facility and operations upgrades (e.g., Project B-534) that were undertaken to enhance the capabilities of the Evaporator. The Evaporator has been classified as a moderate-hazard facility (Johnson 1990). The information contained in this SAR is based on information provided by 242-A Evaporator Operations, Westinghouse Hanford Company, site maintenance and operations contractor from June 1987 to October 1996, and the existing operating contractor, Waste Management Hanford (WMH) policies. Where appropriate, a discussion address the US Department of Energy (DOE) Orders applicable to a topic is provided. Operation of the facility will be compared to the operating contractor procedures using appropriate audits and appraisals. The following subsections provide introductory and background information, including a general description of the Evaporator facility and process, a description of the scope of this SAR revision,a nd a description of the basic changes made to the original SAR

  9. 242-A evaporator safety analysis report

    Energy Technology Data Exchange (ETDEWEB)

    CAMPBELL, T.A.

    1999-05-17

    This report provides a revised safety analysis for the upgraded 242-A Evaporator (the Evaporator). This safety analysis report (SAR) supports the operation of the Evaporator following life extension upgrades and other facility and operations upgrades (e.g., Project B-534) that were undertaken to enhance the capabilities of the Evaporator. The Evaporator has been classified as a moderate-hazard facility (Johnson 1990). The information contained in this SAR is based on information provided by 242-A Evaporator Operations, Westinghouse Hanford Company, site maintenance and operations contractor from June 1987 to October 1996, and the existing operating contractor, Waste Management Hanford (WMH) policies. Where appropriate, a discussion address the US Department of Energy (DOE) Orders applicable to a topic is provided. Operation of the facility will be compared to the operating contractor procedures using appropriate audits and appraisals. The following subsections provide introductory and background information, including a general description of the Evaporator facility and process, a description of the scope of this SAR revision,a nd a description of the basic changes made to the original SAR.

  10. Safety of Research Reactors. Safety Requirements

    International Nuclear Information System (INIS)

    2010-01-01

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

  11. Strategy for seismic upgrading of chemical plant taking productivity as criterion of judgment

    International Nuclear Information System (INIS)

    Oshima, M.; Kase, T.; Yashiro, H.; Fukushima, S.

    2005-01-01

    Seismic upgrading and modification of existing chemical plant facilities have been performed by means of a procedure of the Seismic Design Code and Guidelines of High-pressure Gas Facilities in Japan. Main purpose of this seismic design code is to ensure public safety at seismic events. From the viewpoints of seismic risk of corporate management, CSR (Corporate Social Responsibility) and productivity of the plants are also important for seismic assessment. In this paper, authors proposed strategy for seismic assessment to select appropriate pre-earthquake upgrading and modification considering productivity of plants based on fault tree analysis. This assessment will enable to select weak damage modes and to allocate countermeasure cost optimally to the selected damage modes. (authors)

  12. A human reliability assessment screening method for the NRU upgrade project

    International Nuclear Information System (INIS)

    Bremner, F.M.; Alsop, C.J.

    1997-01-01

    The National Research Universal (NRU) reactor is a 130MW, low pressure, heavy water cooled and moderated research reactor. The reactor is used for research, both in support of Canada's CANDU development program, and for a wide variety of other research applications. In addition, NRU plays an important part in the production of medical isotopes, e.g., generating 80% of worldwide supplies of Molybdenum-99. NRU is owned and operated by Atomic Energy of Canada Ltd. (AECL), and is currently undergoing upgrading as part of AECL's continuing commitment to operate their facilities in a safe manner. As part of these upgrades both deterministic and probabilistic safety assessments are being carried out. It was recognized that the assignment of Human Error Probabilities (HEPs) is an important part of the Probabilistic Safety Assessment (PSA) studies, particularly for a facility whose design predates modern ergonomic practices, and which will undergo a series of backfitted modifications whilst continuing to operate. A simple Human Reliability Assessment (HRA) screening method, looking at both pre- and post-accident errors, was used in the initial safety studies. However, following review of this method within AECL and externally by the regulator, it was judged that benefits could be gained for future error reduction by including additional features, as later described in this document. The HRA development project consisted of several stages; needs analysis, literature review, development of method (including testing and evaluation), and implementation. This paper discusses each of these stages in further detail. (author)

  13. Towards a binding international governance of nuclear safety: an impossible quest?

    International Nuclear Information System (INIS)

    Finon, Dominique

    2014-01-01

    The Fukushima accident again raises the question of the social and economic viability of nuclear technology. On an international basis, it should be necessary to reach the maximum level of safety for every nuclear power plant (NPP) to avoid any further accident in order to preserve the acceptability of the technology. To obtain a significant orientation in upgrading safety standards in matter of NPP design and operation and institutional practices for control and safety in all countries with nuclear facilities, the ideal would be to succeed in setting up a binding international governance. This article examines the incentives and the conditions to achieve it. These incentives on the States appear not to be strong enough at the global level in order that they delegate part of their sovereignty in this domain. It seems that we must be content with a weak governance. This governance combining the role of the IAEA as a facilitator, and different peer pressures mechanisms at the level of the NPP operators, the reactors vendors and the safety authority. We observe that each of these mechanisms is presently being reinforced. But how strong this weak governance is strong enough? (author)

  14. SRS control system upgrade requirements

    International Nuclear Information System (INIS)

    Hill, L.F.

    1998-01-01

    This document defines requirements for an upgrade of the Sodium Removal System (SRS) control system. The upgrade is being performed to solve a number of maintainability and operability issues. The upgraded system will provide the same functions, controls and interlocks as the present system, and in addition provide enhanced functionality in areas discussed in this document

  15. H-1 Upgrades (4BW/4BN) (H-1 Upgrades)

    Science.gov (United States)

    2015-12-01

    automatic blade fold of the new composite rotor blades, new performance matched transmissions, a new four-bladed tail rotor and drive system, upgraded...Upgrades December 2015 SAR March 18, 2016 10:59:17 UNCLASSIFIED 4 Col Steven Girard PMA-276 USMC Light/Attack Helicopter Program Executive Officer...attack helicopter is to provide rotary wing close air support, anti-armor, armed escort, armed/visual reconnaissance and fire support coordination

  16. Selection of fuel design for conversion and upgradation of Pakistan Research Reactor (PARR-1)

    International Nuclear Information System (INIS)

    Arshad, M.

    1991-01-01

    The Pakistan Research Reactor (PARR-1) is being converted from the use of Highly Enriched Uranium (HEU) to Low Enriched Uranium (LEU) fuel and its power is also being upgraded. In order to select new fuel for the converted and upgraded core ten different fuel element designs were analyzed and their relative performance was compared. Results of this study were later used to select appropriate design of the new fuel for PARR-1. This paper describes the computational methodology utilized for the analysis of various fuel element designs. Criteria for selecting the new fuel element are discussed and guidelines forming the selection basis of the new fuel design are given. (author)

  17. Upgrading the ATLAS control system

    International Nuclear Information System (INIS)

    Munson, F.H.; Ferraretto, M.

    1993-01-01

    Heavy-ion accelerators are tools used in the research of nuclear and atomic physics. The ATLAS facility at the Argonne National Laboratory is one such tool. The ATLAS control system serves as the primary operator interface to the accelerator. A project to upgrade the control system is presently in progress. Since this is an upgrade project and not a new installation, it was imperative that the development work proceed without interference to normal operations. An additional criteria for the development work was that the writing of additional ''in-house'' software should be kept to a minimum. This paper briefly describes the control system being upgraded, and explains some of the reasons for the decision to upgrade the control system. Design considerations and goals for the new system are described, and the present status of the upgrade is discussed

  18. LHC luminosity upgrade detector challenges

    CERN Multimedia

    CERN. Geneva; de Roeck, Albert; Bortoletto, Daniela; Wigmans, Richard; Riegler, Werner; Smith, Wesley H

    2006-01-01

    LHC luminosity upgrade: detector challenges The upgrade of the LHC machine towards higher luminosity (1035 cm -2s-1) has been studied over the last few years. These studies have investigated scenarios to achieve the increase in peak luminosity by an order of magnitude, as well as the physics potential of such an upgrade and the impact of a machine upgrade on the LHC DETECTORS. This series of lectures will cover the following topics: • Physics motivation and machine scenarios for an order of magnitude increase in the LHC peak luminosity (lecture 1) • Detector challenges including overview of ideas for R&D programs by the LHC experiments: tracking and calorimetry, other new detector developments (lectures 2-4) • Electronics, trigger and data acquisition challenges (lecture 5) Note: the much more ambitious LHC energy upgrade will not be covered

  19. Technical Barriers, Gaps, and Opportunities Related to Home Energy Upgrade Market Delivery

    Energy Technology Data Exchange (ETDEWEB)

    Bianchi, M. V. A.

    2011-11-01

    This report outlines the technical barriers, gaps, and opportunities that arise in executing home energy upgrade market delivery approaches, as identified through research conducted by the U.S. Department of Energy's Building America program. The objective of this report is to outline the technical1 barriers, gaps, and opportunities that arise in executing home energy upgrade market delivery approaches, as identified through research conducted by the U.S. Department of Energy's (DOE) Building America program. This information will be used to provide guidance for new research necessary to enable the success of the approaches. Investigation for this report was conducted via publications related to home energy upgrade market delivery approaches, and a series of interviews with subject matter experts (contractors, consultants, program managers, manufacturers, trade organization representatives, and real estate agents). These experts specified technical barriers and gaps, and offered suggestions for how the technical community might address them. The potential benefits of home energy upgrades are many and varied: reduced energy use and costs; improved comfort, durability, and safety; increased property value; and job creation. Nevertheless, home energy upgrades do not comprise a large part of the overall home improvement market. Residential energy efficiency is the most complex climate intervention option to deliver because the market failures are many and transaction costs are high (Climate Change Capital 2009). The key reasons that energy efficiency investment is not being delivered are: (1) The opportunity is highly fragmented; and (2) The energy efficiency assets are nonstatus, low-visibility investments that are not properly valued. There are significant barriers to mobilizing the investment in home energy upgrades, including the 'hassle factor' (the time and effort required to identify and secure improvement works), access to financing, and the

  20. Advanced Reactor Safety Program – Stakeholder Interaction and Feedback

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo H. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-08-01

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

  1. The LHCb Muon Upgrade

    CERN Multimedia

    Cardini, A

    2013-01-01

    The LHCb collaboration is currently working on the upgrade of the experiment to allow, after 2018, an efficient data collection while running at an instantaneous luminosity of 2x10$^{33}$/cm$^{-2}$s$^{-1}$. The upgrade will allow 40 MHz detector readout, and events will be selected by means of a very flexible software-based trigger. The muon system will be upgraded in two phases. In the first phase, the off-detector readout electronics will be redesigned to allow complete event readout at 40 MHz. Also, part of the channel logical-ORs, used to reduce the total readout channel count, will be removed to reduce dead-time in critical regions. In a second phase, higher-granularity detectors will replace the ones installed in highly irradiated regions, to guarantee efficient muon system performances in the upgrade data taking conditions.

  2. Development of safety analysis technology for LMR

    International Nuclear Information System (INIS)

    Hahn, Do Hee; Kwon, Y. M.; Suk, S. D.

    2002-05-01

    In the present study, the KALIMER safety analysis has been made for the transients considered in the design concept, hypothetical core disruptive accident (HCDA), and containment performance with the establishment of the design basis. Such analyses have not been possible without the computer code improvement, and the experience attained during this research period must have greatly contributed to the achievement of the self reliance in the domestic technology establishment on the safety analysis areas of the conceptual design. The safety analysis codes have been improved to extend their applicable ranges for detailed conceptual design, and a basic computer code system has been established for HCDA analysis. A code-to-code comparison analysis has been performed as a part of code verification attempt, and the leading edge technology of JNC also has been brought for the technology upgrade. In addition, the research and development on the area of the database establishment has been made for the efficient and systematic project implementation of the conceptual design, through performances on the development of a project scheduling management, integration of the individually developed technology, establishment of the product database, and so on, taking into account coupling of the activities conducted in each specific area

  3. Upgrader alley : oil sands fever strikes Edmonton

    International Nuclear Information System (INIS)

    Griffiths, M.; Dyer, S.

    2008-01-01

    Large-scale industrial complexes called upgraders are similar to oil refineries. Several upgraders are planned for the area just northeast of Edmonton, known as Upgrader Alley. Concerns have been expressed over the potential congestion and environmental impacts of these upgraders. Upgraders will also attract other industry, and the cumulative effects of development will have major impacts on the region, its people and the natural environment. The report provided an overview of Upgrader Alley, with reference to what is driving development; upgrading issues; what Upgrader Alley will look like; and how much water Upgrader Alley needs. The report also discussed impacts on the land, air quality, and greenhouse gases. Water demand issues were discussed with reference to impacts on the North Saskatchewan River, water levels, water quality, a water management framework, and groundwater resources. Cumulative impacts were also presented. It was concluded that if all the projects for which applications had been submitted were approved, the rapid pace of growth in Upgrader Alley would mimic that of Fort McMurray. If the rate of development were somewhat slower, there would be more time to develop and implement plans to reduce the impacts. 189 refs., 6 tabs., 14 figs

  4. Technical basis for environmental qualification of computer-based safety systems in nuclear power plants

    International Nuclear Information System (INIS)

    Korsah, K.; Wood, R.T.; Tanaka, T.J.; Antonescu, C.E.

    1997-01-01

    This paper summarizes the results of research sponsored by the US Nuclear Regulatory Commission (NRC) to provide the technical basis for environmental qualification of computer-based safety equipment in nuclear power plants. This research was conducted by the Oak Ridge National Laboratory (ORNL) and Sandia National Laboratories (SNL). ORNL investigated potential failure modes and vulnerabilities of microprocessor-based technologies to environmental stressors, including electromagnetic/radio-frequency interference, temperature, humidity, and smoke exposure. An experimental digital safety channel (EDSC) was constructed for the tests. SNL performed smoke exposure tests on digital components and circuit boards to determine failure mechanisms and the effect of different packaging techniques on smoke susceptibility. These studies are expected to provide recommendations for environmental qualification of digital safety systems by addressing the following: (1) adequacy of the present preferred test methods for qualification of digital I and C systems; (2) preferred standards; (3) recommended stressors to be included in the qualification process during type testing; (4) resolution of need for accelerated aging in qualification testing for equipment that is to be located in mild environments; and (5) determination of an appropriate approach to address smoke in a qualification program

  5. Upgrading nuclear safeguards in Kazakhstan

    International Nuclear Information System (INIS)

    Hunt, Maribeth; Murakami, Kenji

    2005-01-01

    accurately account for the nuclear material received, and to generally upgrade the safety, security, and accountancy standards. While the material that is actually in hold-up may not be of particular concern with respect to nuclear proliferation and nuclear security, a plant's declaration of hold-up may be a way of concealing diversion of nuclear material. Overstating the amount of material in hold-up can allow an operator to divert material. In the past, neither the UMP nor the IAEA has had an accurate estimate of the material designated as hold-up. When the hold-up of a plant can be characterized and verified it assures that this proliferation pathway is protected. During the PIV in September 2004, the Agency used its ISOCS system to re-measure points that had been measured during the PIV in 2003 and to measure points that the operator had measured with the systems provided by the Japanese and the United States. The result was that at the Physical Inventory Taking in 2004, UMP staff were able to make an effective declaration of hold-up and the Agency was able to verify it. At the same time, the rest of the work of the donor States was progressing. The funding from the Japanese government has assured the preparation of three procedures to standardize the nuclear material accountancy and control at the UMP and has sponsored the training of two UMP safeguards officials in Japan. One of the original goals of this integrated project was to significantly reduce, by 2005, the uncertainty in the measurement of hold-up at the UMP. Through the concerted efforts of the IAEA, donor States, and the European Union, this goal was reached in September 2004. Over the next year, work will concentrate on further training of personnel, the Russian translation and distribution of the procedures funded by the Japanese government, coordinating the delivery and installation of equipment from the United States, and UMP staff training in safeguards and safety culture

  6. Setting priorities for safeguards upgrades

    International Nuclear Information System (INIS)

    Al-Ayat, R.A.; Judd, B.R.; Patenaude, C.J.; Sicherman, A.

    1987-01-01

    This paper describes an analytic approach and a computer program for setting priorities among safeguards upgrades. The approach provides safeguards decision makers with a systematic method for allocating their limited upgrade resources. The priorities are set based on the upgrades cost and their contribution to safeguards effectiveness. Safeguards effectiveness is measured by the probability of defeat for a spectrum of potential insider and outsider adversaries. The computer program, MI$ER, can be used alone or as a companion to ET and SAVI, programs designed to evaluate safeguards effectiveness against insider and outsider threats, respectively. Setting the priority required judgments about the relative importance (threat likelihoods and consequences) of insider and outsider threats. Although these judgments are inherently subjective, MI$ER can analyze the sensitivity of the upgrade priorities to these weights and determine whether or not they are critical to the priority ranking. MI$ER produces tabular and graphical results for comparing benefits and identifying the most cost-effective upgrades for a given expenditure. This framework provides decision makers with an explicit and consistent analysis to support their upgrades decisions and to allocate the safeguards resources in a cost-effective manner

  7. Determination of the design excess reactivity for the TREAT Upgrade reactor

    International Nuclear Information System (INIS)

    Bhattacharyya, S.K.; Hanan, N.A.

    1983-01-01

    The excess reactivity designed to be built into a reactor core is a primary determinant of the fissile loadings of the fuel rods in the core. For the TREAT Upgrade (TU) reactor the considerations that enter into the determination of the excess reactivity are different from those of conventional power reactors. The reactor is designed to operate in an adiabatic transient mode for reactor safety in-pile test programs. The primary constituent of the excess reactivity is the calculated reactivity required to perform the most demanding transient experiments. Because of the unavailability of supporting critical experiments for the core design, the uncertainty terms that add on to this basic constituent are rather large. The burnup effects in TU are negligible and no refueling is planned. In this paper the determination of the design excess reactivity of the TREAT Upgrade reactor is discussed

  8. 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)

  9. Biogas as a renewable energy fuel – A review of biogas upgrading, utilisation and storage

    International Nuclear Information System (INIS)

    Ullah Khan, Imran; Hafiz Dzarfan Othman, Mohd; Hashim, Haslenda; Matsuura, Takeshi; Ismail, A.F.; Rezaei-DashtArzhandi, M.; Wan Azelee, I.

    2017-01-01

    Highlights: • It is attempted to provide critical considerations on various biogas upgrading techniques. • Membrane is an environmentally and economically sound technique for purification. • Purified biogas in compressed form is a substitute of compressed natural gas for vehicles. • Charged mass and compressor input work are the most important factors for storage. - Abstract: Biogas upgrading is a widely studied and discussed topic and its utilisation as a natural gas substitute has gained a significant attention in recent years. The production of biomethane provides a versatile application in both heat and power generation and as a vehicular fuel. This paper systematically reviews the state of the art of biogas upgrading technologies with upgrading efficiency, methane (CH 4 ) loss, environmental effect, development and commercialisation, and challenges in terms of energy consumption and economic assessment. The market situation for biogas upgrading has changed rapidly in recent years, making the membrane separation gets significant market share with traditional biogas upgrading technologies. In addition, the potential utilisation of biogas, efficient conversion into bio-compressed natural gas (bio-CNG), and storage systems are investigated in depth. Two storing systems for bio-CNG at filling stations, namely buffer and cascade storage systems are used. The best storage system should be selected on the basis of the advantages of both systems. Also, the fuel economy and mass emissions for bio-CNG and CNG filled vehicles are studied. There is the same fuel economy and less carbon dioxide (CO 2 ) emission for bio-CNG. Based on the results of comparisons between the technical features of upgrading technologies, various specific requirements for biogas utilisation and the relevant investment, and operating and maintenance costs, future recommendations are made for biogas upgrading.

  10. Microprocessor tester for the treat upgrade reactor trip system

    International Nuclear Information System (INIS)

    Lenkszus, F.R.; Bucher, R.G.

    1984-01-01

    The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. In addition, a programmable Automated Reactor Control System (ARCS) will permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety system is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations

  11. Towards a Formal Basis for Modular Safety Cases

    Science.gov (United States)

    Denney, Ewen; Pai, Ganesh

    2015-01-01

    Safety assurance using argument-based safety cases is an accepted best-practice in many safety-critical sectors. Goal Structuring Notation (GSN), which is widely used for presenting safety arguments graphically, provides a notion of modular arguments to support the goal of incremental certification. Despite the efforts at standardization, GSN remains an informal notation whereas the GSN standard contains appreciable ambiguity especially concerning modular extensions. This, in turn, presents challenges when developing tools and methods to intelligently manipulate modular GSN arguments. This paper develops the elements of a theory of modular safety cases, leveraging our previous work on formalizing GSN arguments. Using example argument structures we highlight some ambiguities arising through the existing guidance, present the intuition underlying the theory, clarify syntax, and address modular arguments, contracts, well-formedness and well-scopedness of modules. Based on this theory, we have a preliminary implementation of modular arguments in our toolset, AdvoCATE.

  12. Bracebridge dam safety presentation

    Energy Technology Data Exchange (ETDEWEB)

    Kulchycky, Vince [Bracebridge Generation Ltd. (Canada)

    2011-07-01

    In Bracebridge, Ontario, hydroelectric generators have been in use since 1901. Bracebridge Generation is now upgrading the Wilson's Falls unit from 600kW to 2900kW. Bracebridge Generation addressed public safety with an independent audit and signage, fencing, booms and buoys installed. Unfortunately these measures did not prevent a recent drowning at the utility's site. More widespread warnings and safety education were seen as measures to stop people from swimming on the utility's property.

  13. Strategy of upgrading management and organizational development of railway transport of Ukraine

    OpenAIRE

    Kompaniec, V.

    2010-01-01

    In article questions of forming of strategy of upgrading management and organizational development are exposed. On the basis of complex analysis the vision is definite: philosophy and case frames, mission, corporate values, complex of strategic targets of the Ukrainian railways, vision of organizational culture development. Also basic aspects are definite of methodology change case frame and organizational culture, including changes in motivation and organizational structure.

  14. Pipeline transportation of emerging partially upgraded bitumen

    International Nuclear Information System (INIS)

    Luhning, R.W.; Anand, A.; Blackmore, T.; Lawson, D.S.

    2002-01-01

    The recoverable reserves of Canada's vast oil deposits is estimated to be 335 billion barrels (bbl), most of which are in the Alberta oil sands. Canada was the largest import supplier of crude oil to the United States in 2001, followed by Saudi Arabia. By 2011, the production of oil sands is expected to increase to 50 per cent of Canada's oil, and conventional oil production will decline as more production will be provided by synthetic light oil and bitumen. This paper lists the announced oil sands projects. If all are to proceed, production would reach 3,445,000 bbl per day by 2011. The three main challenges regarding the transportation and marketing of this new production were described. The first is to expand the physical capacity of existing pipelines. The second is the supply of low viscosity diluent (such as natural gas condensate or synthetic diluent) to reduce the viscosity and density of the bitumen as it passes through the pipelines. The current pipeline specifications and procedures to transport partially upgraded products are presented. The final challenge is the projected refinery market constraint to process the bitumen and synthetic light oil into consumer fuel products. These challenges can be addressed by modifying refineries and increasing Canadian access in Petroleum Administration Defense District (PADD) II and IV. The technology for partial upgrading of bitumen to produce pipeline specification oil, reduce diluent requirements and add sales value, is currently under development. The number of existing refineries to potentially accept partially upgraded product is listed. The partially upgraded bitumen will be in demand for additional upgrading to end user products, and new opportunities will be presented as additional pipeline capacity is made available to transport crude to U.S. markets and overseas. The paper describes the following emerging partial upgrading methods: the OrCrude upgrading process, rapid thermal processing, CPJ process for

  15. The ALICE Inner Tracking System Upgrade

    CERN Document Server

    Siddhanta, Sabyasachi

    2015-01-01

    The long term plan of ALICE (A Large Ion Collider Experiment) at the CERN Large Hadron Collider (LHC) is a detailed investigation and characterisation of the Quark-Gluon Plasma (QGP). ALICE has devised a comprehensive upgrade strategy to enhance its physics capabilities and to exploit the LHC running conditions after the second long shutdown of the LHC scheduled in 2019-20. The upgraded ALICE will focus on high precision measurements of rare probes over a wide range of momenta, which will significantly improve the performance with respect to the present experimental set up. The upgrade strategy is based on the fact that after LS2 LHC will progressively increase its luminosity with Pb beams eventually reaching an interaction rate of about 50 kHz. To exploit the new LHC capabilities, several existing detectors will undergo a substantial upgrade and new detectors will be added. Within this upgrade strategy, the Inner Tracking System (ITS) upgrade forms an important cornerstone, providing precise measurements for...

  16. Software and man-machine interface considerations for a nuclear plant computer replacement and upgrade project

    International Nuclear Information System (INIS)

    Diamond, G.; Robinson, E.

    1984-01-01

    Some of the key software functions and Man-Machine Interface considerations in a computer replacement and upgrade project for a nuclear power plant are described. The project involves the installation of two separate computer systems: an Emergency Response Facilities Computer System (ERFCS) and a Plant Process Computer System (PPCS). These systems employ state-of-the-art computer hardware and software. The ERFCS is a new system intended to provide enhanced functions to meet NRC post-TMI guidelines. The PPCS is intended to replace and upgrade an existing obsolete plant computer system. A general overview of the hardware and software aspects of the replacement and upgrade is presented. The work done to develop the upgraded Man-Machine Interface is described. For the ERFCS, a detailed discussion is presented of the work done to develop logic to evaluate the readiness and performance of safety systems and their supporting functions. The Man-Machine Interface considerations of reporting readiness and performance to the operator are discussed. Finally, the considerations involved in the implementation of this logic in real-time software are discussed.. For the PPCS, a detailed discussion is presented of some new features

  17. Maintenance procedure upgrade programs

    International Nuclear Information System (INIS)

    Campbell, J.J.; Zimmerman, C.M.

    1988-01-01

    This paper describes a systematic approach to upgrading nuclear power plant maintenance procedures. The approach consists of four phases: diagnosis, program planning, program implementation, and program evaluation. Each phase is explained as a series of steps to ensure that all factors in a procedure upgrade program are considered

  18. Seismic safety of nuclear power plants in Eastern Europe

    International Nuclear Information System (INIS)

    Gurpinar, A.; Godoy, A.

    1995-01-01

    This paper summarizes the work performed by the International Atomic Energy Agency in the areas of safety reviews and applied research in support of programmes for the assessment and enhancement of seismic safety in WWER type nuclear power plants during the past five years. Three major topics are discussed; engineering safety review services in relation to external events, technical guidelines for the assessment and upgrading of WWER type nuclear power plants, and the Coordinated Research Programme on B enchmark study for the seismic analysis and testing of WWER type nuclear power plants . These topics are summarized in a way to provide an overview of the past and present safety situation in selected WWER type plants which are all located in Eastern European countries. Main conclusion of the paper is that although there is now a thorough understanding of the seismic safety issues in these operating nuclear power plants, the implementation of seismic upgrades to structures, systems and components are lagging behind, particularly for those cases in which the re-evaluation indicated the necessity to strengthen the safety related structures or install new safety systems. (author)

  19. The deep geologic repository technology programme: toward a geoscience basis for understanding repository safety

    International Nuclear Information System (INIS)

    Jensen, M.R.

    2007-01-01

    Within the Deep Geologic Repository Technology Programme (DGRTP) several Geoscience activities are focused on advancing the understanding of groundwater flow system evolution and geochemical stability in a Canadian Shield setting as affected by long-term climate change. A key aspect is developing confidence in predictions of groundwater flow patterns and residence times as they relate to the safety of a deep geologic repository for used nuclear fuel waste. This is being achieved through a coordinated multi-disciplinary approach intent on: i) demonstrating coincidence between independent geo-scientific data; ii) improving the traceability of geo-scientific data and its interpretation within a conceptual descriptive model(s); iii) improving upon methods to assess and demonstrate robustness in flow domain prediction(s) given inherent flow domain uncertainties (i.e. spatial chemical/physical property distributions, boundary conditions) in time and space; and iv) improving awareness amongst geo-scientists as to the utility of various geo-scientific data in supporting a safety case for a deep geologic repository. This multi-disciplinary DGRTP approach is yielding an improved understanding of groundwater flow system evolution and stability in Canadian Shield settings that is further contributing to the geo-scientific basis for understanding and communicating aspects of DGR safety. (author)

  20. Design concepts for the reactor protection and control process instrumentation digital upgrade project at the Donald C. Cook Nuclear Plant units 1 and 2

    International Nuclear Information System (INIS)

    Carruth, R.C.; Sotos, W.G.

    1996-01-01

    As the nation's nuclear power plants age, the need to consider upgrading of their electronic protection and control systems becomes more urgent. Hardware obsolescence and mechanical wear out resulting from frequent calibration and surveillance play a major role in defining their useful life. At Cook Nuclear Plant, a decision was made to replace a major portion of the plant's protection and control systems with newer technology. This paper describes the engineering processes involved in this successful upgrade and explains the basis for many decisions made while performing the digital upgrade

  1. Successful Completion of the Top-off Upgrade of the Advanced Light Source

    International Nuclear Information System (INIS)

    Steier, C.; Bailey, B.; Baptiste, K.; Barry, W.; Biocca, A.; Byrne, W.; Casey, P.; Chin, M.; Donahue, R.; Duarte, R.; Fahmie, M.; Gath, B.; Jacobson, S.; Julian, J.; Jung, J.Y.; Kritscher, M.; Kwiatkowski, S.; Marks, S.; McKean, P.; Mueller, R.; Nishimura, H.; ONeill, J.; Portmann, G.; Prestemon, S.; Robin, D.; Rossi, S.; Sannibale, F.; Scarvie, T.; Schlueter, R.; Shuman, D.; Smith-Baumann, A.; Stover, G.; Timossi, C.; Wan, W.; Warwick, T.; Wells, R.; Weber, J.; Williams, E.

    2010-01-01

    An upgrade of the Advanced Light Source to enable top-off operation has been completed during the last four years. The final work centered around radiation safety aspects, culminating in a systematic proof that top-off operation is equally safe as decaying beam operation. Commissioning and transition to full user operations happened in late 2008 and early 2009. Top-off operation at the ALS provides a very large increase in time-averaged brightness (by about a factor of 10) as well as improvements in beam stability. The following sections provide an overview of the radiation safety rationale, commissioning results, as well as experience in user operations.

  2. Los Alamos controlled air incinerator upgrade for TRU/mixed waste operations

    International Nuclear Information System (INIS)

    Vavruska, J.S.; Borduin, L.C.; Hutchins, D.A.; Warner, C.L.; Thompson, T.K.

    1989-01-01

    The Los Alamos Controlled Air Incinerator (CAI) is undergoing a major process upgrade to accept Laboratory-generated transuranic (TRU) and TRU mixed wastes on a production basis. In the interim,prior to the scheduled 1992 operation of a new on-site LLW/mixed waste incinerator, the CAI will also be accepting solid and liquid low-level mixed wastes. This paper describes major modifications that have been made to the process to enhance safety and ensure reliability for long-term, routine waste incineration operations. The regulatory requirements leading to operational status of the system are also briefly described. The CAI was developed in the mid-1970s as a demonstration system for volume reduction of TRU combustible solid wastes. It continues as a successful R and D system well into the 1980s during which incineration tests on a wide variety of radioactive and chemical waste forms were performed. In 1985, a DOE directive required Los Alamos to reduce the volume of its TRU waste prior to ultimate placement in the geological repository at the Waste Isolation Pilot Project (WIPP). With only minor modifications to the original process flowsheet, the Los Alamos CAI was judged capable of conversion to a TRU waste operations mode. 9 refs., 1 fig

  3. Completion of the VVER 440/213 NPP Mochovce incorporation enhanced safety features

    International Nuclear Information System (INIS)

    Charbonneau, S.; Eckert, G.

    1996-01-01

    The cooperation between the western countries and the countries of ex-eastern block in the field of nuclear safety is recent and still limited. The main reasons for this situation are limited or non existent capabilities of these countries for financing as well as non acceptable legal conditions concerning the third party nuclear liability in this part of Europe. Nevertheless, Framatome and Siemens associated in the consortium named EUCOM, have signed in April 1996 the contract of about 100 million US dollars with Slovak electricity company (SLOVENSKE ELEKTRARNE-SE) for upgrading the Units 1 and 2 of Mochovce Nuclear Power Plant according to the western safety standards. This is the first important project involving west-european companies in the modernisation of Russian type of pressurized water reactor (VVER 440/213). The consortium will cooperate with other partners involved in the project: Slovak, Czech and Russian. The financing of the project will be provided mainly form Slovak and Czech sources. The safety upgrading will be financed through French and German buyer credits. French company Electricite de France (EDF) will be the consultant for SE. The safety upgrading measures have been elaborated taking into account the recommendation of Vienna International Atomic Energy Agency (IAEA) and the evaluation of the safety realised by RISKAUDIT, the common organization of German and French safety authorities (GSR and IPSN). Hence all guaranties have been taken to fulfil the western safety criteria for Nuclear Power Plant Mochovce. (author)

  4. Bracebridge dam safety presentation

    Energy Technology Data Exchange (ETDEWEB)

    Kulchycky, Vince [Bracebridge Generation Ltd. (Canada)

    2011-07-01

    In Bracebridge, Ontario, hydroelectric generators have been in use since 1901. Bracebridge Generation is now upgrading the Wilson's Falls unit from 600kW to 2900kW. Bracebridge Generation addressed public safety with an independent audit and signage, fencing, booms and buoys installed. Unfortunately these measures did not prevent a recent drowning at the utility's site. More widespread warnings and safety education were seen as measures to stop people from swimming on the utility's property.

  5. Probabilistic safety assessment activities at Ignalina NPP

    International Nuclear Information System (INIS)

    Bagdonas, A.

    1999-01-01

    The Barselina Project was initiated in the summer 1991. The project was a multilateral co-operation between Lithuania, Russia and Sweden up until phase 3, and phase 4 has been performed as a bilateral between Lithuania and Sweden. The long-range objective is to establish common perspectives and unified bases for assessment of severe accident risks and needs for remedial measures for the RBMK reactors. During phase 3, from 1993 to 1994, a full scope Probabilistic Safety Analysis (PSA) model of the Ignalina Nuclear Power Plant unit 2 was developed to identify possible safety improvement of risk importance. The probabilistic methodology was applied on a plant specific basis for a channel type reactor of RBMK design. During phase 4, from 1994 to 1996, the PSA was further developed, taking into account plant changes, improved modelling methods and extended plant information concerning dependencies (area events, dynamic effects, electrical and signal dependencies). The model reflected the plant status before the outage 1996. During phase 4+, 1998 to 1999 the PSA model was upgraded taking into account the newest plant modifications. The new PSA model of CPS/AZRT was developed. Modelling was based on the Single Failure Analysis

  6. The biological basis of plutonium safety standards

    International Nuclear Information System (INIS)

    Mole, R.H.

    1976-01-01

    Since no radiation injury or cancer in man can, as yet, be directly attributed to Pu, all safety standards for Pu must be determined by reference to other safety standards, development of which is discussed. A system of safety standards must be based on links with real damage, such as the requirement for 226 Ra in bone. The type of biological information required for making standards realistic is considered in relation to Pu and Ra in bone. Also considered are the possible effects of Pu in soft tissue such as bone marrow. Not only dose, but also the number of cells exposed to the dose are important biologically and cellular aspects are examined. Since there is no positive evidence of Pu toxicity relevant information on other α emitters must be examined. The observed effectiveness of Ra, daughters of 222 Ra and 232 Th in causing mutations and cancer, is surveyed. Reference is made to the necessity of improving the ICRP system, currently based on the critical organ concept, by recognising the need for summation of risks in other organs where exposure to Pu is concerned. Improved biological understanding particularly that of hereditary damage, in recent years leads to less pessimistic thinking on the effects of ionizing radiations. The immediate need appears to be for consistency in safety standards. (U.K.)

  7. Upgrades of the ATLAS trigger system

    CERN Document Server

    AUTHOR|(INSPIRE)INSPIRE-00221618; The ATLAS collaboration

    2018-01-01

    In coming years the LHC is expected to undergo upgrades to increase both the energy of proton-proton collisions and the instantaneous luminosity. In order to cope with these more challenging LHC conditions, upgrades of the ATLAS trigger system will be required. This talk will focus on some of the key aspects of these upgrades. Firstly, the upgrade period between 2019-2021 will see an increase in instantaneous luminosity to $3\\times10^{34} \\rm{cm^{-2}s^{-1}}$. Upgrades to the Level 1 trigger system during this time will include improvements for both the muon and calorimeter triggers. These include the upgrade of the first-level Endcap Muon trigger, the calorimeter trigger electronics and the addition of new calorimeter feature extractor hardware, such as the Global Feature Extractor (gFEX). An overview will be given on the design and development status the aforementioned systems, along with the latest testing and validation results. \\\\ By 2026, the High Luminosity LHC will be able to deliver 14 TeV collisions ...

  8. Alberta oil sands crudes : upgrading and marketing

    International Nuclear Information System (INIS)

    Ashar, M.

    2008-01-01

    Open pit mining and in situ techniques, such as steam stimulation, are used to recover Alberta's bitumen and heavy oil resources, which have higher viscosities than conventional hydrocarbons. The bitumen is typically upgraded to synthetic crude oil (SCO). In the simplest processing scheme, the bitumen is blended with diluent for ease in pipeline transport and then processed at refineries with upgrading facilities. The bitumen is also upgraded to light SCO at world-scale upgraders in Alberta. The SCO is then processed at refineries in downstream markets. The 2 categories of upgrading, notably primary and secondary upgrading, were described in this article along with technology options for both categories. Slurry hydrocracking is regarded as the most interesting emerging residual fuel upgrading technology. It combines special catalyst mixes with the latest slurry reactor designs as well as innovative catalyst capture and recycle schemes to produce very high conversions and potentially superior upgrading economics. The increase in volume and rate of SCO from Alberta provides refiners in the oil sands marketing sector an unprecedented choice of opportunities to improve profitability. Key trends indicate that production will increase substantially from 2008 to 2030. 5 figs

  9. Alberta oil sands crudes : upgrading and marketing

    Energy Technology Data Exchange (ETDEWEB)

    Ashar, M. [Suncor Energy, Fort McMurray, AB (Canada)

    2008-05-15

    Open pit mining and in situ techniques, such as steam stimulation, are used to recover Alberta's bitumen and heavy oil resources, which have higher viscosities than conventional hydrocarbons. The bitumen is typically upgraded to synthetic crude oil (SCO). In the simplest processing scheme, the bitumen is blended with diluent for ease in pipeline transport and then processed at refineries with upgrading facilities. The bitumen is also upgraded to light SCO at world-scale upgraders in Alberta. The SCO is then processed at refineries in downstream markets. The 2 categories of upgrading, notably primary and secondary upgrading, were described in this article along with technology options for both categories. Slurry hydrocracking is regarded as the most interesting emerging residual fuel upgrading technology. It combines special catalyst mixes with the latest slurry reactor designs as well as innovative catalyst capture and recycle schemes to produce very high conversions and potentially superior upgrading economics. The increase in volume and rate of SCO from Alberta provides refiners in the oil sands marketing sector an unprecedented choice of opportunities to improve profitability. Key trends indicate that production will increase substantially from 2008 to 2030. 5 figs.

  10. Examination of alternatives to upgrade fire safety in operating NPPs

    International Nuclear Information System (INIS)

    Park, J. H.; Jung, I. S.

    2003-01-01

    For the Kori unit 2,3,4, Younggwang 1,2 and Ulchin 1,2 fire protection vulnerabilities and further improvement items are investigated. The most major problem is that those plants do not have plant-specific Fire Hazard Analysis Report. To improve fire safety of those plants, their own Fire Hazard Analysis should be proceed at first. Then, according to results of Fire Hazard Analysis, fire protection improvement program should be implemented. For the assurance of long term fire safety, result of Fire Hazard Analysis should be incorporated in the FSAR and periodically reviewed the impact of design change to fire safety

  11. Supporting Fernald Site Closure with Integrated Health and Safety Plans as Documented Safety Analyses

    International Nuclear Information System (INIS)

    Kohler, S.; Brown, T.; Fisk, P.; Krach, F.; Klein, B.

    2004-01-01

    At the Fernald Closure Project (FCP) near Cincinnati, Ohio, environmental restoration activities are supported by Documented Safety Analyses (DSAs) that combine the required project-specific Health and Safety Plans, Safety Basis Requirements (SBRs), and Process Requirements (PRs) into single Integrated Health and Safety Plans (I-HASPs). These integrated DSAs employ Integrated Safety Management methodology in support of simplified restoration and remediation activities that, so far, have resulted in the decontamination and demolition (D and D) of over 200 structures, including eight major nuclear production plants. There is one of twelve nuclear facilities still remaining (Silos containing uranium ore residues) with its own safety basis documentation. This paper presents the status of the FCP's safety basis documentation program, illustrating that all of the former nuclear facilities and activities have now replaced. Basis of Interim Operations (BIOs) with I-HASPs as their safety basis during the closure process

  12. The Radiation Safety Interlock System for Top-Up Mode Operation at NSRRC

    CERN Document Server

    Chen Chien Rong; Kao, Sheau-Ping; Liu, Joseph; Sheu, Rong-Jiun; Wang, Jau-Ping

    2005-01-01

    The radiation safety interlock systems of NSRRC have been operated for more than a decade. Some modification actions have been implemented in the past to perfect the safe operation. The machine and its interlock system were originally designed to operate at the decay mode. Recently some improvement programs to make the machine injection from original decay mode to top-up mode at NSRRC has initiated. For users at experimental area the radiation dose resulted from top-up re-fill injections where safety shutters of beam-lines are opened will dominate. In addition to radiation safety action plans such as upgrading the shielding, enlarging the exclusion zones and improving the injection efficiency, the interlock system for top-up operation is the most important to make sure that injection efficiency is acceptable. To ensure the personnel radiation safety during the top-up mode, the safety interlock upgrade and action plans will be implemented. This paper will summarize the original design logic of the safety inter...

  13. To upgrade or not to upgrade

    DEFF Research Database (Denmark)

    Rasmussen, Torben Valdbjørn; Rose, Jørgen

    2017-01-01

    This paper identifies the key indicators that owners need to take into account in order to choose the most affordable extent of upgrading of a typical post-1945 building. The exterior look of the building is not to be changed. Indicators include measures, risk assessment related to the changes in...

  14. AGS intensity upgrades

    International Nuclear Information System (INIS)

    Roser, T.

    1995-01-01

    After the successful completion of the AGS Booster and several upgrades of the AGS, a new intensity record of 6.3 x 10 13 protons per pulse accelerated to 24 GeV was achieved. The high intensity slow-extracted beam program at the AGS typically serves about five production targets and about eight experiments including three rare Kaon decay experiments. Further intensity upgrades are being discussed that could increase the average delivered beam intensity by up to a factor of four

  15. Upgrading uncompetitive products economically

    DEFF Research Database (Denmark)

    Lu, Hua; Jensen, Christian Søndergaard

    2012-01-01

    for upgrading an uncompetitive product, and combine the solutions into a single solution. We also propose a spatial join-based solution that assumes P and T are indexed by an R-tree. Given a set of products in the same R-tree node, we derive three lower bounds on their upgrading costs. These bounds are employed...

  16. Developing Major Steps for a Feasibility Study for Upgrading I and C Systems in a Large Scale for an Operating Nuclear Power Plant

    International Nuclear Information System (INIS)

    Suh, Yong Suk; Keum, Jong Yong; Kim, Dong Hoon; Kang, Hyeon Tae; Sung, Chan Ho; Lee, Jae Ki; Cho, Chang Hwan

    2009-01-01

    According to the IAEA report as of Jan. 2008, 436 nuclear power reactors are in operation over the world and 368 nuclear power reactors exceed their operating ages by 20 years. The average I and C equipment's life span is 20 years comparing with that the average reactor's life time is 40 to 60 years. This means that a reactor must be faced with I and C equipment obsolescence problems once or twice during its operating years. The I and C equipment is replaced with new equipment only when the obsolescence problem occurs in a nuclear power plant. This is called an equipment basis upgrade in this paper. This replacement is such a general practice that occurs only when needed. We can assume that most of I and C equipment of a plant will meet with the obsolescence problem almost same time since it started operating. Although there must be a little time difference in the occurrence of the problems among I and C equipment, the replacement will be required in consecutive years. With this assumption, it is recommendable to upgrade the equipment, which is to meet with the problem at the same time, with new equipment at the same time. This is called a system basis upgrade in this paper. The system-basis replacement can be achieved in a large scale by coupling systems whose functions are related each other and replacing them together with a new upto- date platform. This paper focuses on the large scale upgrade of I and C systems for existing and operating NPPs. While performing a feasibility study for the large scale upgrade for Korea standard nuclear power plants (KSNPs), six major steps are developed for the study. This paper is to present what to perform in each step

  17. Kinematic optimization of upgrade to the Hobby-Eberly Telescope through novel use of commercially available three-dimensional CAD package

    Science.gov (United States)

    Wedeking, Gregory A.; Zierer, Joseph J.; Jackson, John R.

    2010-07-01

    The University of Texas, Center for Electromechanics (UT-CEM) is making a major upgrade to the robotic tracking system on the Hobby Eberly Telescope (HET) as part of theWide Field Upgrade (WFU). The upgrade focuses on a seven-fold increase in payload and necessitated a complete redesign of all tracker supporting structure and motion control systems, including the tracker bridge, ten drive systems, carriage frames, a hexapod, and many other subsystems. The cost and sensitivity of the scientific payload, coupled with the tracker system mass increase, necessitated major upgrades to personnel and hardware safety systems. To optimize kinematic design of the entire tracker, UT-CEM developed novel uses of constraints and drivers to interface with a commercially available CAD package (SolidWorks). For example, to optimize volume usage and minimize obscuration, the CAD software was exercised to accurately determine tracker/hexapod operational space needed to meet science requirements. To verify hexapod controller models, actuator travel requirements were graphically measured and compared to well defined equations of motion for Stewart platforms. To ensure critical hardware safety during various failure modes, UT-CEM engineers developed Visual Basic drivers to interface with the CAD software and quickly tabulate distance measurements between critical pieces of optical hardware and adjacent components for thousands of possible hexapod configurations. These advances and techniques, applicable to any challenging robotic system design, are documented and describe new ways to use commercially available software tools to more clearly define hardware requirements and help insure safe operation.

  18. 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.

  19. Captain upgrade CRM training: A new focus for enhanced flight operations

    Science.gov (United States)

    Taggart, William R.

    1993-01-01

    Crew Resource Management (CRM) research has resulted in numerous payoffs of applied applications in flight training and standardization of air carrier flight operations. This paper describes one example of how basic research into human factors and crew performance was used to create a specific training intervention for upgrading new captains for a major United States air carrier. The basis for the training is examined along with some of the specific training methods used, and several unexpeced results.

  20. Lessons learned from the safety assistance program for soviet-designed reactors

    International Nuclear Information System (INIS)

    Steinberg, N.

    1999-01-01

    Two examples of nuclear power situation were compared in this conference paper - the situation in Lithuania and the situation in the Ukraine. Based on the examples mentioned, author conclude that the effectiveness of the Multi-National Safety Assistance Program for Soviet -Designed Reactors in a given recipient country does not depend, in practice, on engineering issues. The principal aspects that determine this effectiveness are: first, the level of safety culture in the country, beginning at the Governmental level but also at the level of the senior managers of nuclear power. The other important factor which contributes is the availability of a well-developed national program for upgrading NPP safety. The economical well-being of nuclear power and of the country as a whole also has a major effect on the effectiveness of the western technical assistance programs that are trying to upgrade reactor safety in a particular recipient country. And finally, international community should have well coordinated and well substantiated safety assistance program for specific country

  1. The ALICE TPC Upgrad

    Science.gov (United States)

    Castro, Andrew; Alice-Usa Collaboration; Alice-Tpc Collaboration

    2017-09-01

    The Time Projection Chamber (TPC) currently used for ALICE (A Large Ion Collider Experiment at CERN) is a gaseous tracking detector used to study both proton-proton and heavy-ion collisions at the Large Hadron Collider (LHC) In order to accommodate the higher luminosit collisions planned for the LHC Run-3 starting in 2021, the ALICE-TPC will undergo a major upgrade during the next LHC shut down. The TPC is limited to a read out of 1000 Hz in minimum bias events due to the intrinsic dead time associated with back ion flow in the multi wire proportional chambers (MWPC) in the TPC. The TPC upgrade will handle the increase in event readout to 50 kHz for heavy ion minimum bias triggered events expected with the Run-3 luminosity by switching the MWPCs to a stack of four Gaseous Electron Multiplier (GEM) foils. The GEM layers will combine different hole pitches to reduce the dead time while maintaining the current spatial and energy resolution of the existing TPC. Undertaking the upgrade of the TPC represents a massive endeavor in terms of design, production, construction, quality assurance, and installation, thus the upgrade is coordinated over a number of institutes worldwide. The talk will go over the physics motivation for the upgrade, the ALICE-USA contribution to the construction of Inner Read Out Chambers IROCs, and QA from the first chambers built in the U.S

  2. ATLAS Upgrade Programme

    CERN Document Server

    Hillier, S J; The ATLAS collaboration

    2012-01-01

    With the already outstanding LHC luminosity performance, and planned LHC upgrades in the upcoming shutdowns, it is expected that within a short time-scale, the general purpose LHC experiments will have to cope with luminosities beyond their original design. In order to maintain detector performance and sensitivity to expected and new physics processes, ATLAS has defined a continuous upgrade programme which foresees staged enhancements during the next 10 years of operation, and then more widespread changes before the transition to the highest luminosities after 2022. This talk will describe several components of the ATLAS upgrade, focusing in particular on the Inner Detector and Trigger. The Inner Detector faces two challenges in the higher luminosity environment: high particle multiplicities and increased radiation dose. These will be addressed in the short term by a new layer of Pixel detectors, and in the long term by a complete replacement. The Trigger faces an increasingly difficult task of distinguishing...

  3. Selection of design basis event for modular high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Nakagawa, Shigeaki; Ohashi, Hirofumi

    2016-06-01

    Japan Atomic Energy Agency (JAEA) has been investigating safety requirements and basic approach of safety guidelines for modular High Temperature Gas-cooled Reactor (HTGR) aiming to increase internarial contribution for nuclear safety by developing an international HTGR safety standard under International Atomic Energy Agency. In this study, we investigate a deterministic approach to select design basis events utilizing information obtained from probabilistic approach. In addition, selections of design basis events are conducted for commercial HTGR designed by JAEA. As a result, an approach for selecting design basis event considering multiple failures of safety systems is established which has not been considered as design basis in the safety guideline for existing nuclear facility. Furthermore, selection of design basis events for commercial HTGR has completed. This report provides an approach and procedure for selecting design basis events of modular HTGR as well as selected events for the commercial HTGR, GTHTR300. (author)

  4. Imaging motional Stark effect measurements at ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Ford, O. P.; Burckhart, A.; McDermott, R.; Pütterich, T.; Wolf, R. C. [Max-Planck Institut für Plasmaphysik, Greifswald/Garching (Germany)

    2016-11-15

    This paper presents an overview of results from the Imaging Motional Stark Effect (IMSE) diagnostic obtained during its first measurement campaign at ASDEX Upgrade since installation as a permanent diagnostic. A brief overview of the IMSE technique is given, followed by measurements of a standard H-mode discharge, which are compared to equilibrium reconstructions showing good agreement where expected. The development of special discharges for the calibration of pitch angle is reported and safety factor profile changes during sawteeth crashes are shown, which can be resolved to a few percent due to the high sensitivity at good time resolution of the new IMSE system.

  5. Turbulence in high-beta ASDEX upgrade advanced scenarios

    Science.gov (United States)

    Doerk, H.; Bock, A.; Di Siena, A.; Fable, E.; Görler, T.; Jenko, F.; Stober, J.; The ASDEX Upgrade Team

    2018-01-01

    Recent experiments at ASDEX Upgrade achieve non-inductive operation in full tungsten wall conditions by applying electron cyclotron and neutral beam current drive. These discharges are characterised by a well-measured safety factor profile, which does not drop below one, and a good energy confinement. By reproducing the experimental heat fluxes, nonlinear gyrokinetic simulations suggest that the observed strong peaking of the ion temperature in the core is caused by the stabilising impact of a significant beam ion content, as well as strong electromagnetic effects on turbulent transport. Quasilinear transport models are not yet applicable in this interesting and reactor relevant parameter regime, but available simulation data may serve as a testbed for improvements. As the present plasma is close to the kinetic ballooning (KBM) threshold, elevating the safety factor profile under otherwise identical conditions is proposed to clarify, whether profiles are ultimately limited by KBM turbulence, or by global stability constraints.

  6. Seismic safety of Paks nuclear power plant

    International Nuclear Information System (INIS)

    Katona, T.

    1995-01-01

    This paper contains an overview of the results concerning the following activities: investigation of methods, regulations and techniques for reassessment of seismic safety of operating NPPs and upgrading of safety; investigation of earthquake hazards; development of concept for creation of the seismic safety location of earthquake warning system; determination of dynamic features of systems and facilities determined by the concept and preliminary evaluation of the seismic safety. It is limited on investigation of dynamic features of building structures, the building dynamical experiments and experimental investigation of the equipment

  7. Upgrade of the AGS H- linac

    International Nuclear Information System (INIS)

    Alessi, J.G.; Buxton, W.; Kponou, A.; LoDestro, V.; Mapes, M.; McNerney, A.J.; Raparia, D.

    1994-01-01

    The AGS linac presently accelerates 25 mA of H - to 200 MeV at a 5 Hz rep-rate and 500 μs pulse width. The Booster takes 4 pulses every 3.8 seconds, and the remaining pulses are used for isotope production. The authors are in the process of upgrading the linac to increase the average current delivered for isotope production by more than a factor of two, while at the same time expecting to decrease linac downtime. Various aspects of this upgrade are discussed, including the upgrade of the control system, new high power transmission line, transport line vacuum, and rf power supply system upgrades

  8. Implementation of U.S. Department of Energy physical protection upgrades in Lithuania and Uzbekistan

    International Nuclear Information System (INIS)

    Haase, M.; Romesberg, L.; Showalter, R.; Soo Hoo, M.S.; Corey, J.; Engling, E.

    1996-01-01

    Since 1994, the U.S. Department of Energy (DOE) has provided cooperative assistance to the non-nuclear weapons states of the Former Soviet Union. This effort, within DOE's program of Material Protection, Control, and Accounting (MPC ampersand A), identified the Institute of Nuclear Physics (INP) in Uzbekistan and the Ignalina Nuclear Power Plant (INPP) in Lithuania as sites for cooperative MPC ampersand A projects. The INP, located just outside of Tashkent, is the site of a 10-megawatt WWR-SM research reactor. This reactor is expected to remain operational as a major nuclear research and isotope production reactor for Central Asia. The INPP, located 100 kilometers northeast of the capital city of Vilnius, consists of two Russian-made RBMK reactors with a combined power output of 3,000 megawatts (electric). This power plant has been the subject of international safety and security concerns, which prompted DOE's cooperative assistance effort. This paper describes U.S. progress in a multi-national effort directed at implementing physical protection upgrades in Lithuania and Uzbekistan. The upgrades agreed upon between DOE and the INP and between DOE and the INPP have been designed to interface with upgrades being implemented by other donor countries. DOE/INPP upgrade projects include providing training on U.S. approaches to physical protection, access control through the main vehicle portal, a hardened central alarm station, and improved guard force communications. DOE/INP upgrade projects in Uzbekistan include an access control system, a hardened fresh fuel storage vault, an interior intrusion detection and assessment system, and an integrated alarm display and assessment system

  9. TECHNICAL BASIS DOCUMENT FOR NATURAL EVENT HAZARDS

    International Nuclear Information System (INIS)

    KRIPPS, L.J.

    2006-01-01

    This technical basis document was developed to support the documented safety analysis (DSA) and describes the risk binning process and the technical basis for assigning risk bins for natural event hazard (NEH)-initiated accidents. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous conditions based on an evaluation of the frequency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR-level controls

  10. Biogas Upgrading and Waste-to-Energy | Bioenergy | NREL

    Science.gov (United States)

    Biogas Upgrading and Waste-to-Energy Biogas Upgrading and Waste-to-Energy NREL's waste-to-energy research and development required for upgrading biogas to fuels and high-value co-products. Featured (2014) Biogas Potential in the United States, NREL Fact Sheet (2013) View all NREL biogas upgrading and

  11. The Safety Case and Safety Assessment for the Disposal of Radioactive Waste

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-09-15

    This Safety Guide provides guidance and recommendations on meeting the safety requirements in respect of the safety case and supporting safety assessment for the disposal of radioactive waste. The safety case and supporting safety assessment provide the basis for demonstration of safety and for licensing of radioactive waste disposal facilities and assist and guide decisions on siting, design and operations. The safety case is also the main basis on which dialogue with interested parties is conducted and on which confidence in the safety of the disposal facility is developed. This Safety Guide is relevant for operating organizations preparing the safety case as well as for the regulatory body responsible for developing the regulations and regulatory guidance that determine the basis and scope of the safety case. Contents: 1. Introduction; 2. Demonstrating the safety of radioactive waste disposal; 3. Safety principles and safety requirements; 4. The safety case for disposal of radioactive waste; 5. Radiological impact assessment for the period after closure; 6. Specific issues; 7. Documentation and use of the safety case; 8. Regulatory review process.

  12. Staying Secure for School Safety

    Science.gov (United States)

    Youngkin, Minu

    2012-01-01

    Proper planning and preventive maintenance can increase school security and return on investment. Preventive maintenance begins with planning. Through careful planning, education institutions can determine what is working and if any equipment, hardware or software needs to be replaced or upgraded. When reviewing a school's safety and security…

  13. The ASDEX Upgrade ICRF system: Operational experience and developments

    International Nuclear Information System (INIS)

    Faugel, H.; Angene, P.; Becker, W.; Braun, F.; Bobkov, Vl.V.; Eckert, B.; Fischer, F.; Hartmann, D.A.; Heilmaier, G.; Kneidl, J.; Noterdaeme, J.-M.; Siegl, G.; Wuersching, E.

    2005-01-01

    The ICRF system at the ASDEX Upgrade tokamak is in operation since May 1992. Following some modifications of which the major one was the installation of 3 dB couplers it has become a reliable additional heating system. The maximum power coupled into the plasma has been raised up to 7.2 MW (90% of the installed RF power) for short pulses and up to 6.2 MW for pulses several second long with energy of up to 29 MJ. A power of 5 MW is delivered on a regular basis to replace two NBI sources

  14. Five TWh saved on profitable upgrading

    International Nuclear Information System (INIS)

    Bakken, Stein Arne.

    1999-01-01

    Article. In Norway, a project is planned to evaluate the upgrading potential of the hydroelectric power sector. The article criticizes the authorities for not supporting the upgrading of hydroelectric power plants in their eager to promote the development of wind power. An important contribution of the planned project will be a book from which the utilities may get advice on measures of upgrading. Generators and control systems account for 80% of the faults and interruptions in the power utilities. The turbine technology has made important progress in the 1990s. By pure upgrading measures alone, the turbine efficiencies may be increased enough for an additional 3 TWh to be taken out from the Norwegian hydroelectric power system. Today, the price of imported carbon-based Danish power determines whether the utilities find it profitable to rehabilitate the hydroelectric plants. The energy potential of upgrading is estimated to be 5 TWh. This energy can be generated with known technology and with no serious consequences for the environment

  15. Upgrading the ATLAS Tile Calorimeter electronics

    CERN Document Server

    Carrio, F; The ATLAS collaboration

    2013-01-01

    The Tile Calorimeter (TileCal) is the hadronic calorimeter covering the most central region of the ATLAS experiment at LHC. The TileCal readout consists of about 10000 channels. Its main upgrade will occur for the High Luminosity LHC phase (phase 2) where the luminosity will have increased 5-fold compared to the design luminosity (1034 cm−2s−1) but with maintained energy (i.e. 7+7 TeV). An additional luminosity increase by a factor of 2 can be achieved by luminosity leveling. This upgrade will probably happen around 2022. The upgrade aims at replacing the majority of the on- and off- detector electronics so that all calorimeter signals are directly digitized and sent to the off-detector electronics in the counting room. To achieve the required reliability, redundancy has been introduced at different levels. An ambitious upgrade development program is pursued studying different electronics options. Three different options are presently being investigated for the front-end electronic upgrade. Which one to u...

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

    International Nuclear Information System (INIS)

    Batandjieva, B.; Torres-Vidal, C.

    2002-01-01

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

  17. Ferrocyanide Safety Program: Safety criteria for ferrocyanide watch list tanks

    International Nuclear Information System (INIS)

    Postma, A.K.; Meacham, J.E.; Barney, G.S.

    1994-01-01

    This report provides a technical basis for closing the ferrocyanide Unreviewed Safety Question (USQ) at the Hanford Site. Three work efforts were performed in developing this technical basis. The efforts described herein are: 1. The formulation of criteria for ranking the relative safety of waste in each ferrocyanide tank. 2. The current classification of tanks into safety categories by comparing available information on tank contents with the safety criteria; 3. The identification of additional information required to resolve the ferrocyanide safety issue

  18. Heavy water upgrading system in the Fugen heavy water reactor

    International Nuclear Information System (INIS)

    Matsushita, T.; Susaki, S.

    1980-01-01

    The heavy water upgrading system, which is installed in the Fugen heavy water reactor (HWR) was designed to reuse degraded heavy water generated from the deuteration-dedeuteration of resin in the ion exchange column of the moderator purification system. The electrolysis method has been applied in this system on the basis of the predicted generation rate and concentration of degraded heavy water. The structural feature of the electrolytic cell is that it consists of dual cylindrical electrodes, instead of a diaphragm as in the case of conventional water electrolysis. 2 refs

  19. Upgrade trigger: Biannual performance update

    CERN Document Server

    Aaij, Roel; Couturier, Ben; Esen, Sevda; De Cian, Michel; De Vries, Jacco Andreas; Dziurda, Agnieszka; Fitzpatrick, Conor; Fontana, Marianna; Grillo, Lucia; Hasse, Christoph; Jones, Christopher Rob; Le Gac, Renaud; Matev, Rosen; Neufeld, Niko; Nikodem, Thomas; Polci, Francesco; Del Buono, Luigi; Quagliani, Renato; Schwemmer, Rainer; Seyfert, Paul; Stahl, Sascha; Szumlak, Tomasz; Vesterinen, Mika Anton; Wanczyk, Joanna; Williams, Mark Richard James; Yin, Hang; Zacharjasz, Emilia Anna

    2017-01-01

    This document presents the performance of the LHCb Upgrade trigger reconstruction sequence, incorporating changes to the underlying reconstruction algorithms and detector description since the Trigger and Online Upgrade TDR. An updated extrapolation is presented using the most recent example of an Event Filter Farm node.

  20. Upgrading the safety toolkit: Initiatives of the accident analysis subgroup

    International Nuclear Information System (INIS)

    O'Kula, K.R.; Chung, D.Y.

    1999-01-01

    Since its inception, the Accident Analysis Subgroup (AAS) of the Energy Facility Contractors Group (EFCOG) has been a leading organization promoting development and application of appropriate methodologies for safety analysis of US Department of Energy (DOE) installations. The AAS, one of seven chartered by the EFCOG Safety Analysis Working Group, has performed an oversight function and provided direction to several technical groups. These efforts have been instrumental toward formal evaluation of computer models, improving the pedigree on high-use computer models, and development of the user-friendly Accident Analysis Guidebook (AAG). All of these improvements have improved the analytical toolkit for best complying with DOE orders and standards shaping safety analysis reports (SARs) and related documentation. Major support for these objectives has been through DOE/DP-45

  1. Light Water Reactor Sustainability Program Technical Basis Guide Describing How to Perform Safety Margin Configuration Risk Management

    Energy Technology Data Exchange (ETDEWEB)

    Curtis Smith; James Knudsen; Bentley Harwood

    2013-08-01

    The INL has carried out a demonstration of the RISMC approach for the purpose of configuration risk management. We have shown how improved accuracy and realism can be achieved by simulating changes in risk – as a function of different configurations – in order to determine safety margins as the plant is modified. We described the various technical issues that play a role in these configuration-based calculations with the intent that future applications can take advantage of the analysis benefits while avoiding some of the technical pitfalls that are found for these types of calculations. Specific recommendations have been provided on a variety of topics aimed at improving the safety margin analysis and strengthening the technical basis behind the analysis process.

  2. Processing options for bitumen upgrading

    International Nuclear Information System (INIS)

    Harji, A.N.; Koppel, P.E.; Mazurek, W.L.; Meysami, P.

    2003-01-01

    It is estimated that 178 billion barrels of oil can be recovered from Alberta's vast heavy oil reserves. The challenge lies in the logistics of recovering, upgrading and transporting the oil to market. The Canadian Energy Research Institute conducted a recent study to determine market potential by 2007 for diluted bitumen and synthetic crude oil produced from upgraded bitumen. The viability for a wide range of bitumen feedstocks was assessed along with the sensitivity of refinery demand to their prices. The 3 major markets for western Canadian bitumen include PADD 2 in the United States, western Canada, and Ontario. Bitumen is too viscous to transport by pipeline and cannot be processed by most of the existing refineries. Therefore, in order to develop a mass market for the product, bitumen must undergo the energy intensive upgrading process at existing refineries. The factors impacting which method of upgrading is most suitable were discussed with particular attention to the impact that Canada's ratification of the Kyoto Protocol may have on Alberta's bitumen resource in terms of costs of complying with greenhouse gas reduction initiatives. The authors emphasized that it is crucial to customize an upgrading project to meet site and market specific factors. 8 refs., 3 tabs., 3 figs

  3. Technical basis for evaluating electromagnetic and radio-frequency interference in safety-related I ampersand C systems

    International Nuclear Information System (INIS)

    Ewing, P.D.; Korsah, K.

    1994-04-01

    This report discusses the development of the technical basis for the control of upsets and malfunctions in safety-related instrumentation and control (I ampersand C) systems caused by electromagnetic and radio-frequency interference (EMI/RFI) and power surges. The research was performed at the Oak Ridge National Laboratory (ORNL) and was sponsored by the USNRC Office of Nuclear Regulatory Research (RES). The motivation for research stems from the safety-related issues that need to be addressed with the application of advanced I ampersand C systems to nuclear power plants. Development of the technical basis centered around establishing good engineering practices to ensure that sufficient levels of electromagnetic compatibility (EMC) are maintained between the nuclear power plant's electronic and electromechanical systems known to be the source(s) of EMI/RFI and power surges. First, good EMC design and installation practices need to be established to control the impact of interference sources on nearby circuits and systems. These EMC good practices include circuit layouts, terminations, filtering, grounding, bonding, shielding, and adequate physical separation. Second, an EMI/RFI test and evaluation program needs to be established to outline the tests to be performed, the associated test methods to be followed, and carefully formulated acceptance criteria based on the intended environment to ensure that the circuit or system under test meets the recommended guidelines. Third, a program needs to be developed to perform confirmatory tests and evaluate the surge withstand capability (SWC) and of I ampersand C equipment connected to or installed in the vicinity of power circuits within the nuclear power plant. By following these three steps, the design and operability of safety-related I ampersand C systems against EMI/RFI and power surges can be evaluated, acceptance criteria can be developed, and appropriate regulatory guidance can be provided

  4. Review of selected cost drivers for decisions on continued operation of older nuclear reactors. Safety upgrades, lifetime extension, decommissioning

    International Nuclear Information System (INIS)

    1999-05-01

    Lately, the approach to the operation of relatively old NPPs has become an important issue for the nuclear industry for several reasons. First, a large part of operating NPPs will reach the planned end of their lives relatively soon. Replacing these capacities can involve significant investment for the concerned countries and utilities. Second, many operating NPPs while about 30 years old are still in very good condition. Their continued safe operation appears possible and may bring about essential economic gains. Finally, with the costs of new NPPs being rather high at present, continued operation of existing plants and eventually their lifetime extension are viable options for supporting the nuclear share in power generation. This is becoming especially important in view of the growing attention to the issue of global warming and the role of nuclear energy in greenhouse gas mitigation. This report is a review of information related to three cost categories that are part of such cost-benefit analysis: costs of safety upgrades for continued operation of a nuclear unit, costs of lifetime extension and costs of decommissioning. It can serve as a useful reference source for experts and decision makers involved in the economics of operating NPPs

  5. Pulsed power supply for Nova Upgrade

    International Nuclear Information System (INIS)

    Bacon, J.L.; Kajs, J.P.; Walls, A.; Weldon, W.F.; Zowarka, R.C.

    1992-01-01

    This report describes work carried out at the Center for Electromechanics at The University of Texas at Austin (CEM-UT). A baseline design of the Nova Upgrade has been completed by Lawrence Livermore National Laboratory. The Nova Upgrade is an 18 beamline Nd: glass laser design utilizing fully relayed 4x4 30 cm aperture segmented optical components. The laser thus consists of 288 independent beamlets nominally producing 1.5 to 2.0 MJ of 0.35 μm light in a 3 to 5 ns pulse. The laser design is extremely flexible and will allow a wide range of pulses to irradiate ICF targets. This facility will demonstrate ignition/gain and the scientific feasibility of ICF for energy and defense applications. The pulsed power requirements for the Nova Upgrade are given. CEM-UT was contracted to study and develop a design for a homopolar generator/inductor (HPG/inductor) opening switch system which would satisfy the pulsed power supply requirements of the Nova Upgrade. The Nd:glass laser amplifiers used in the Nova Upgrade will be powered by light from xenon flashlamps. The pulsed power supply for the Nova Upgrade powers the xenon flashlamps. This design and study was for a power supply to drive flashlamps

  6. Hydrogen alternatives for a regional upgrader

    International Nuclear Information System (INIS)

    Bailey, R.T.; Padamsey, R.

    1991-01-01

    For a proposed regional upgrader in Alberta, hydrogen will be needed to upgrade the bitumen and heavy oil to be processed by that facility. The upgrader will rely on high conversion hydrocracking which consumes 3.4 wt % hydrogen to produce a 106% volume yield of high quality synthetic crude. The costs of producing hydrogen via steam reforming of methane, partial oxidation of coal or upgrading residues, and electrolysis are compared, showing that steam reforming is the cheapest. However, an even cheaper source of hydrogen is available in the Edmonton and Fort Saskatchewan area as byproducts from petrochemical plants. An economic analysis is presented of a proposed scheme to capture, purify, compress, and transfer hydrogen from one or two such plants to a nearby regional upgrader. The two plants could supply a total of 126.6 million ft 3 /d of hydrogen at a total installed capital cost of about half of that of a steam reforming plant of equivalent size. When operating costs are added (including the cost of replacing the hydrogen, currently used as fuel at the two plants, with natural gas), the total cost of hydrogen is substantially less than the costs for a hydrogen plant within the upgrader. 3 refs., 5 figs., 4 tabs

  7. Safety analysis for research reactors

    International Nuclear Information System (INIS)

    2008-01-01

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

  8. Safety assessment as basis for the decision making process

    International Nuclear Information System (INIS)

    Ilie, P.; Didita, L.; Danchiv, A.

    2005-01-01

    This paper deals with the safety assessment for a new near surface repository, particularly for the early stage of repository development using ISAM (Improvement of Safety Assessment Methodologies for Near Surface Disposal Facilities) safety assessment methodology. In this stage of the repository life cycle the main purpose of the safety assessment is to demonstrate that the plant is capable to be constructed and operated safely. The paper is based on development of the ASAM (Application of the Safety Assessment Methodologies for Near-Surface Disposal Facilities) Decision Support Subgroup of the Common Aspects Working Group. The implications of decision making for the application of the ISAM methodology on post-closure safety assessment are analysed. Some important elements of the decision-making process with impact on key components of the ISAM process are described. Following the development of Decision Support Subgroup of the ASAM Common Aspects Working Group the proposed change of ISAM methodology is analysed. This approach puts all activities in a decision context where the first iteration of the safety assessment is based on the existing state of knowledge and the initial engineering design. Confidence in the process is accomplished through the direct inclusion of all decision makers and stakeholders in the formulation of decisions, the definition of the state of knowledge, and decision making activities. The decision process is developed in context of undertaking assessments with little site-specific information, this situation is specifically for new planned repository. Limited site-specific information can result in a high degree of uncertainty, therefore it is important first of all to identify the sources of uncertainty arising from the limited nature of the site-specific information and then to apply appropriate approaches to manage the uncertainties and to determine whether the uncertainties are important to the overall safety of the disposal facility

  9. Combustion Safety Simplified Test Protocol Field Study

    Energy Technology Data Exchange (ETDEWEB)

    Brand, L. [Gas Technology Inst., Des Plaines, IL (United States); Cautley, D. [Gas Technology Inst., Des Plaines, IL (United States); Bohac, D. [Gas Technology Inst., Des Plaines, IL (United States); Francisco, P. [Gas Technology Inst., Des Plaines, IL (United States); Shen, L. [Gas Technology Inst., Des Plaines, IL (United States); Gloss, S. [Gas Technology Inst., Des Plaines, IL (United States)

    2015-11-01

    Combustions safety is an important step in the process of upgrading homes for energy efficiency. There are several approaches used by field practitioners, but researchers have indicated that the test procedures in use are complex to implement and provide too many false positives. Field failures often mean that the house is not upgraded until after remediation or not at all, if not include in the program. In this report the PARR and NorthernSTAR DOE Building America Teams provide a simplified test procedure that is easier to implement and should produce fewer false positives. A survey of state weatherization agencies on combustion safety issues, details of a field data collection instrumentation package, summary of data collected over seven months, data analysis and results are included. The project team collected field data on 11 houses in 2015.

  10. LHCb VELO Upgrade

    CERN Document Server

    Hennessy, Karol

    2016-01-01

    The upgrade of the LHCb experiment, scheduled for LHC Run-III, scheduled to start in 2021, will transform the experiment to a trigger-less system reading out the full detector at 40 MHz event rate. All data reduction algorithms will be executed in a high-level software farm enabling the detector to run at luminosities of $2\\times10^{33} \\mathrm{cm}^{-2}\\mathrm{s}^{-1}$. The Vertex Locator (VELO) is the silicon vertex detector surrounding the interaction region. The current detector will be replaced with a hybrid pixel system equipped with electronics capable of reading out at 40 MHz. The upgraded VELO will provide fast pattern recognition and track reconstruction to the software trigger. The silicon pixel sensors have $55\\times55 \\mu m^{2}$ pitch, and are read out by the VeloPix ASIC, from the Timepix/Medipix family. The hottest region will have pixel hit rates of 900 Mhits/s yielding a total data rate of more than 3 Tbit/s for the upgraded VELO. The detector modules are located in a separate vacuum, separate...

  11. ATLAS Upgrade Plans

    CERN Document Server

    Hopkins, W; The ATLAS collaboration

    2014-01-01

    After the successful LHC operation at the center-of-mass energies of 7 and 8 TeV in 2010-2012, plans are actively advancing for a series of upgrades of the accelerator, culminating roughly ten years from now in the high-luminosity LHC (HL-LHC) project, delivering of the order of five times the LHC nominal instantaneous luminosity along with luminosity leveling. The final goal is to extend the dataset from about few hundred fb−1 expected for LHC running to 3000/fb by around 2035 for ATLAS and CMS. In parallel, the experiments need to be keep lockstep with the accelerator to accommodate running beyond the nominal luminosity this decade. Current planning in ATLAS envisions significant upgrades to the detector during the consolidation of the LHC to reach full LHC energy and further upgrades. The challenge of coping with the HL-LHC instantaneous and integrated luminosity, along with the associated radiation levels, requires further major changes to the ATLAS detector. The designs are developing rapidly for a new...

  12. Upgrading of Sergiev Posad department of Moscow NPO Radon

    International Nuclear Information System (INIS)

    Debieve, Pierre; Delecaut, Gregory; Vanleeuw, Daniel

    2007-01-01

    Available in abstract form only. Full text of publication follows: BELGATOM and IRE Consortium has been awarded by the European Commission end of 2005 to conduct a project entitled 'Upgrading of Sergiev Posad Department of Moscow NPO Radon and the assessment of the radiological impact in the area nearby'. The main aims to achieve in the frame of this Europe-aid Project are: - Improvement of the performance and the safety level of the present radwaste management system, taking into account the additional waste expected from the Kurchatov Institute rehabilitation and from the forecast decommissioning of Research Reactors on the territory of Moscow. - Basic design and assistance for the procurement of upgrading equipment related to: - radwaste sorting and pretreatment - replacement of the hydraulic system of the existing super-compactor - characterisation system for radwaste 'Support for preparing the PSAR and PEIAR for new licensing' Assessment of the radiological impact in an area of 50 km radius around Sergiev Posad Department. - The initial duration of this Project is 3 years, starting beginning of 2006. This paper describes the difficulties encountered to start and implement the Project and its status at the half of the planned time schedule. (authors)

  13. Lessons learned in digital upgrade projects digital control system implementation at US nuclear power stations

    International Nuclear Information System (INIS)

    Kelley, S.; Bolian, T. W.

    2006-01-01

    AREVA NP has gained significant experience during the past five years in digital upgrades at operating nuclear power stations in the US. Plants are seeking modernization with digital technology to address obsolescence, spare parts availability, vendor support, increasing age-related failures and diminished reliability. New systems offer improved reliability and functionality, and decreased maintenance requirements. Significant lessons learned have been identified relating to the areas of licensing, equipment qualification, software quality assurance and other topics specific to digital controls. Digital control systems have been installed in non safety-related control applications at many utilities within the last 15 years. There have also been a few replacements of small safety-related systems with digital technology. Digital control systems are proving to be reliable, accurate, and easy to maintain. Digital technology is gaining acceptance and momentum with both utilities and regulatory agencies based upon the successes of these installations. Also, new plants are being designed with integrated digital control systems. To support plant life extension and address obsolescence of critical components, utilities are beginning to install digital technology for primary safety-system replacement. AREVA NP analyzed operating experience and lessons learned from its own digital upgrade projects as well as industry-wide experience to identify key issues that should be considered when implementing digital controls in nuclear power stations

  14. LHCb VELO upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Hennessy, Karol

    2017-02-11

    The upgrade of the LHCb experiment, scheduled for LHC Run-III, scheduled to start in 2021, will transform the experiment to a trigger-less system reading out the full detector at 40 MHz event rate. All data reduction algorithms will be executed in a high-level software farm enabling the detector to run at luminosities of 2×10{sup 33} cm{sup −2} s{sup −1}. The Vertex Locator (VELO) is the silicon vertex detector surrounding the interaction region. The current detector will be replaced with a hybrid pixel system equipped with electronics capable of reading out at 40 MHz. The upgraded VELO will provide fast pattern recognition and track reconstruction to the software trigger. The silicon pixel sensors have 55×55 μm{sup 2} pitch, and are read out by the VeloPix ASIC, from the Timepix/Medipix family. The hottest region will have pixel hit rates of 900 Mhits/s yielding a total data rate of more than 3 Tbit/s for the upgraded VELO. The detector modules are located in a separate vacuum, separated from the beam vacuum by a thin custom made foil. The foil will be manufactured through milling and possibly thinned further by chemical etching. The material budget will be minimised by the use of evaporative CO{sub 2} coolant circulating in microchannels within 400 μm thick silicon substrates. The current status of the VELO upgrade is described and latest results from operation of irradiated sensor assemblies are presented.

  15. Experience in the upgrading of radioactive waste disposal facility 'Ekores'

    International Nuclear Information System (INIS)

    Rozdyalovskaya, L.

    2000-01-01

    The national Belarus radioactive disposal facility 'Ekores' is designed for waste from nuclear applications in industry, medicine and research. Currently 12-20 tons of waste and over 6000 various types spent sources annually come to the 'Ekores'. Total activity in the vaults is evaluated as 352.8 TBq. Approximately 150 000 spent sources disposed of in the vaults and wells have total activity about 1327 TBq. In 1997 the Government initiated a project for the facility reconstruction in order to upgrade radiological safety of the site by creating adequate safety conditions for managing and storage of the waste. The reconstruction project developed by Belarus specialists has been reviewed by IAEA experts. This covers modernising technologies for new coming waste and also that the waste currently disposed in the pits is retrieved, sorted and treated in the same way as the new coming waste

  16. GTHTR300 cost reduction through design upgrade and cogeneration

    Energy Technology Data Exchange (ETDEWEB)

    Yan, Xing L., E-mail: yan.xing@jaea.go.jp; Sato, Hiroyuki; Kamiji, Yu; Imai, Yoshiyuki; Terada, Atsuhiko; Tachibana, Yukio; Kunitomi, Kazuhiko

    2016-09-15

    Japan Atomic Energy Agency began design and development of the Gas Turbine High Temperature Reactor of 300MWe nominal output (GTHTR300) in 2001. The reactor baseline design completed three years later was based on 850 °C core outlet temperature and a direct cycle gas turbine balance of plant. It attained 45.6% net power generation efficiency and 3.5 US¢/kW h cost of electricity. The cost was estimated 20% lower than LWR. The latest design upgrade has incorporated several major technological advances made in the past ten years to both reactor and balance of plant. As described in this paper, these advances have enabled raising the design basis reactor core outlet temperature to 950 °C and increasing power generating efficiency by nearly 5% point. Further implementation of seawater desalination cogeneration is made through employing a newly-proposed multi-stage flash process. Through efficient waste heat recovery of the reactor gas turbine power conversion cycle, a large cost credit is obtained against the conventionally produced water prices. Together, the design upgrade and the cogeneration are shown to reduce the GTHTR300 cost of electricity to under 2.7 US¢/kW h.

  17. GTHTR300 cost reduction through design upgrade and cogeneration

    International Nuclear Information System (INIS)

    Yan, Xing L.; Sato, Hiroyuki; Kamiji, Yu; Imai, Yoshiyuki; Terada, Atsuhiko; Tachibana, Yukio; Kunitomi, Kazuhiko

    2014-01-01

    Japan Atomic Energy Agency began design and development of the Gas Turbine High Temperature Reactor of 300MWe nominal output (GTHTR300) in 2001. The reactor baseline design completed three years later was based on 850°C core outlet temperature and a direct cycle gas turbine balance of plant. It attained 45.6% net power generation efficiency and 3.5US¢/KWh cost of electricity. The cost was estimated 20% lower than LWR. The latest design upgrade has incorporated several major technological advances made in the past ten years to both reactor and balance of plant. As described in this paper, these advances have enabled raising the design basis reactor core outlet temperature to 950°C and increasing power generating efficiency by nearly 5% point. Further implementation of seawater desalination cogeneration is made through employing a newly-proposed multi-stage flash process. Through efficient waste heat recovery of the reactor gas turbine power conversion cycle, a large cost credit is obtained against the conventionally produced water prices. Together, the design upgrade and the cogeneration are shown to reduce the GTHTR300 cost of electricity to under 2.7 US¢/KWh. (author)

  18. OMEGA Upgrade preliminary design

    International Nuclear Information System (INIS)

    Craxton, R.S.

    1989-10-01

    The OMEGA laser system at the Laboratory for Laser Energetics of the University of Rochester is the only major facility in the United States capable of conducting fully diagnosed, direct-drive, spherical implosion experiments. As such, it serves as the national Laser Users Facility, benefiting scientists throughout the country. The University's participation in the National Inertial Confinement Fusion (ICF) program underwent review by a group of experts under the auspices of the National Academy of Sciences (the Happer Committee) in 1985. The Happer Committee recommended that the OMEGA laser be upgraded in energy to 30 kJ. To this end, Congress appropriated $4,000,000 for the preliminary design of the OMEGA Upgrade, spread across FY88 and FY89. This document describes the preliminary design of the OMEGA Upgrade. The proposed enhancements to the existing OMEGA facility will result in a 30-kHJ, 351-nm, 60-beam direct-drive system, with a versatile pulse-shaping facility and a 1%--2% uniformity of target drive. The Upgrade will allow scientists to explore the ignition-scaling regime, and to study target behavior that is hydrodynamically equivalent to that of targets appropriate for a laboratory microfusion facility (LMF). In addition, it will be possible to perform critical interaction experiments with large-scale-length uniformly irradiated plasmas

  19. Common basis of establishing safety standards and other safety decision-making levels for different sources of health risk

    International Nuclear Information System (INIS)

    Demin, V.F.

    2002-01-01

    Current approaches in establishing safety standards and other decision-making levels for different sources of health risk are critically analysed. To have a common basis for this decision-making a specific risk index R is recommended. In the common sense R is quantitatively defined as LLE caused by the annual exposure to the risk source considered: R = annual exposure, damage (LLE) from the exposure unit. This common definition is also rewritten in specific forms for a set of different risk sources (ionising radiation, chemical pollutants, etc): for different risk sources the exposure can be measured with different quantities (the probability of death, the exposure dose, etc.). R is relative LLE: LLE in years referred to 1 year under the risk. The dimension of this value is [year/year]. In the statistical sense R is conditionally the share of the year, which is lost due to exposure to a risk source during this year. In this sense R can be called as the relative damage. Really lifetime years are lost after the exposure. R can be in some conditional sense considered as a dimensionless quantity. General safety standards R n for the public and occupational workers have been suggested in terms of this index: R n = 0.0007 and 0.01 accordingly. Secondary safety standards are derived for a number of risk sources (ionising radiation, environmental chemical pollutants, etc). Values of R n are chosen in such a way that to have the secondary radiation BSS being equivalent to the current one's. Other general and derived levels for safety decision-making are also proposed including the de-minimus levels. Their possible dependence on the national or regional health-demographic data (HDD) is considered. Such issues as the ways of the integration and averaging of risk indices considered through the national or regional HDD for different risk sources and the use of non-threshold linear exposure - response relationships for ionising radiation and chemical pollutants are analysed

  20. Technical basis document for natural event hazards

    International Nuclear Information System (INIS)

    CARSON, D.M.

    2003-01-01

    This technical basis document was developed to support the Tank Farms Documented Safety Analysis (DSA), and describes the risk binning process and the technical basis for assigning risk bins for natural event hazards (NEH)-initiated representative accident and associated represented hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous conditions based on an evaluation of the frequency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR-level controls. Determination of the need for safety-class SSCs was performed in accordance with DOE-STD-3009-94, ''Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'', as described in this report

  1. Superconducting magnet development for the LHC upgrades

    International Nuclear Information System (INIS)

    Rossi, Lucio

    2012-01-01

    LHC is now delivering proton and heavy ion collisions at the highest energy. Upgrading the LHC beyond its design performance is a long term program that started during the LHC construction, with some fundamental R and D programs. The upgrade program is based on a vigorous superconductor and magnet R and D, aimed at increasing the field in accelerator magnets from 8 T to 12 T for the luminosity upgrade, with the scope of increasing the collider luminosity by a factor 5 to 10 from 2022. The upgrade program might continue with the LHC energy upgrade, which would require magnets producing field in the range of 16-20 T. The results obtained so far and the future challenges are discussed together with the possible plan to reach the goals. (author)

  2. Status of safety analysis reports

    Energy Technology Data Exchange (ETDEWEB)

    Cserhati, A

    1999-06-01

    The safety regulation connected to both of the Atomic Acts from 1980 and 1996 requires preparation of the Preliminary Safety Analysis Report (PSAR) as well as Final SAR (FSAR). In this respect the licensing procedure for the construction and commissioning of Paks NPP did not formally deviate from the standards applied in developed countries; this is particularly true if comparison is made with the standards applied for commissioning NPPs in the second half of the seventies. By the time the overall development of internationally accepted safety standards and some existing deficiencies of earlier SAR made necessary a general reassessment of the plant safety (AGNES project). The carried out PSR for Paks-1 and 2 also added a valuable contribution to the SAR content, however a formal update of SAR is not made yet. A Hungarian nuclear authority decree from 1997 obligates the licensee to prepare and submit a major upgrade of FSAR until the mid of 2000, after finishing the PSR for Paks-3 and 4. From this date a periodic update of FSAR is required every year. The operational license renewal affects only the PSR but not the FSAR updating. The new Nuclear Safety Code outlines the contents of PSAR and FSAR, based on US NRC Reg. Guide 1. 70. Rev. 3. Hungary by now can fulfill the upgrading of SAR without major external technical or financial help. The AGNES project covered the safety analysis chapters of SAR. It was financed mainly by the country. In the project there have been involved in limited cases as performers the VTT (Finland), Belgatom (Belgium), GRS (Germany), etc., the IVO (Finland) fulfilled tasks of an independent reviewer for safety analysis. The AGNES had certain interconnection with the similar IAEA RER safety reassessment project for WWER-440/213. The PSR for Paks-1 and 2 have been carried out by the Paks staff from the resources of the plant. During the evaluation of several parts of Paks-3 and 4 PSR documentation the authority intends to use certain

  3. Status of safety analysis reports

    International Nuclear Information System (INIS)

    Cserhati, A.

    1999-01-01

    The safety regulation connected to both of the Atomic Acts from 1980 and 1996 requires preparation of the Preliminary Safety Analysis Report (PSAR) as well as Final SAR (FSAR). In this respect the licensing procedure for the construction and commissioning of Paks NPP did not formally deviate from the standards applied in developed countries; this is particularly true if comparison is made with the standards applied for commissioning NPPs in the second half of the seventies. By the time the overall development of internationally accepted safety standards and some existing deficiencies of earlier SAR made necessary a general reassessment of the plant safety (AGNES project). The carried out PSR for Paks-1 and 2 also added a valuable contribution to the SAR content, however a formal update of SAR is not made yet. A Hungarian nuclear authority decree from 1997 obligates the licensee to prepare and submit a major upgrade of FSAR until the mid of 2000, after finishing the PSR for Paks-3 and 4. From this date a periodic update of FSAR is required every year. The operational license renewal affects only the PSR but not the FSAR updating. The new Nuclear Safety Code outlines the contents of PSAR and FSAR, based on US NRC Reg. Guide 1. 70. Rev. 3. Hungary by now can fulfill the upgrading of SAR without major external technical or financial help. The AGNES project covered the safety analysis chapters of SAR. It was financed mainly by the country. In the project there have been involved in limited cases as performers the VTT (Finland), Belgatom (Belgium), GRS (Germany), etc., the IVO (Finland) fulfilled tasks of an independent reviewer for safety analysis. The AGNES had certain interconnection with the similar IAEA RER safety reassessment project for WWER-440/213. The PSR for Paks-1 and 2 have been carried out by the Paks staff from the resources of the plant. During the evaluation of several parts of Paks-3 and 4 PSR documentation the authority intends to use certain

  4. Assessment of the long-term safety of repositories. Scientific basis

    International Nuclear Information System (INIS)

    Noseck, Ulrich; Becker, Dirk; Fahrenholz, Christine

    2008-12-01

    The project contributed to increase the scientific knowledge on the long-term safety assessment and the safety cases of a radioactive waste repository. International guidelines and more recent safety cases from other countries were evaluated. The feasibility study of the three safety indicators ''individual dose rate'', ''radiotoxicity concentration in the biosphere water'' and ''radiotoxicity flux from the geosphere'' showed that due to the independently derived corresponding reference values these indicators describe three different safety statements. The combination of the three values can give a stronger argument for the safety of the repository system. Another important methodological aspect of the safety cases is the definition and selection of scenarios, one of these the human intrusion scenario. Various human intrusion scenarios are considered in the different nations, which differ significantly with respect to type and time scale, the exposition type and exposition pathway. Further progress has been achieved in how to treat human intrusion scenarios in a German post-closure safety case. Another port of the project dealt with the impact of specific geochemical processes on the long-term safety of the repository. The impact of climate changes on the long-term safety of a radioactive waste repository in rock salt was investigated with respect to processes in the overburden and the biosphere where highest impact is expected. Sofa simplified models and only discrete climate estates have been considered

  5. Completion of the brightness upgrade of the ALS

    International Nuclear Information System (INIS)

    Steier, C; Madur, A; Bailey, B; Berg, K; Biocca, A; Black, A; Casey, P; Colomb, D; Gunion, B; Li, N; Marks, S; Nishimura, H; Pappas, C; Petermann, K; Portmann, G; Prestemon, S; Rawlins, A; Robin, D; Rossi, S; Scarvie, T

    2014-01-01

    The Advanced Light Source (ALS) at Berkeley Lab remains one of the brightest sources for soft x-rays worldwide. A multiyear upgrade of the ALS is underway, which includes new and replacement x-ray beamlines, a replacement of many of the original insertion devices and many upgrades to the accelerator. The accelerator upgrade that affects the ALS performance most directly is the ALS brightness upgrade [1], which reduces the horizontal emittance from 6.3 to 2.0 nm (2.5 nm effective). Magnets for this upgrade were installed in late 2012 and early 2013 followed by user operation with the reduced emittance.

  6. UPGRADES

    CERN Multimedia

    D. Contardo and J. Spalding

    2012-01-01

      Good progress is being made on the projects that will be installed during LS1. CSC chamber production for ME4/2 is progressing at a rate of four chambers per month, with 25 built so far, and the new electronics for ME1/1 is undergoing a pre-production integration testing. For the RPC chambers, gap production is underway with first deliveries to the chamber assembly sites at CERN and Ghent. The third site at Mumbai will begin production next month. For the PMT replacement in the forward hadron calorimeters (HF), the 1728 PMTs are all characterised and ready to be installed. Testing of the electronics boards is going well. Preparations to replace the HPDs in the outer calorimeter (HO) with SiPMs are also on-track. All components are at CERN and burn-in of the new front-end electronics is proceeding. There are three major upgrade projects targeting the period from LS1 through LS2: a new pixel detector, upgraded photo-detectors and electronics for HCAL, and development of a new L1 Trigger. The new ...

  7. UPGRADES

    CERN Multimedia

    Didier Contardo

    2012-01-01

      The CMS Upgrade Programme is making good progress on the LS1 and Phase 1 projects, in the planning for Phase 2. The construction of the ME4/2 muon chambers to be installed during LS1 has started and the two first CSC production chambers have been fully qualified. The three muon groups have recently established a set of milestones towards the completion of their project, that will be integrated in the detailed planning and scheduling for the shutdown work established by Technical Coordination. The project to replace the photo-detectors in the HF and HO calorimeters is also well advanced and at the validation stage. The operation of an HF slice with new multi-anode PMTs and back-end electronics has already been demonstrated in 2012. For the Phase 1 data-taking, as discussed in the Chamonix workshop, it is likely that the LHC performance will exceed the nominal luminosity and pile-up before the second shutdown, still scheduled in 2018. The collaboration is therefore pursuing a strategy to upgrade ...

  8. UPGRADES

    CERN Multimedia

    D. Contardo and J. Spalding

    2013-01-01

    The three post-LS1 Phase 1 Upgrade projects (the L1-Trigger, Pixel Tracker, and HCAL) are all making excellent progress and are transitioning from the prototype to the execution phase. Meanwhile plans are developing for Phase 2, a major Upgrade programme targeting the third long shutdown, LS3. News on Phase 1 is included under the respective projects; we only provide a brief summary here. Phase 1 The plan for the L1 Trigger relies on the installation during the present shutdown of optical splitting for the Trigger input signals. This will allow the new Trigger system to be brought online and fully commissioned during beam operation in 2015, while CMS relies on the existing legacy Trigger for physics. Once fully commissioned the experiment can switch over to the new Trigger, which will provide greatly improved performance at high event pile-up, by 2016. System tests of the splitter system, and of the new architecture of the calorimeter trigger were very successful, and the work in LS1 is on-track. Prototype ...

  9. Using system dynamics simulation for assessment of hydropower system safety

    Science.gov (United States)

    King, L. M.; Simonovic, S. P.; Hartford, D. N. D.

    2017-08-01

    Hydropower infrastructure systems are complex, high consequence structures which must be operated safely to avoid catastrophic impacts to human life, the environment, and the economy. Dam safety practitioners must have an in-depth understanding of how these systems function under various operating conditions in order to ensure the appropriate measures are taken to reduce system vulnerability. Simulation of system operating conditions allows modelers to investigate system performance from the beginning of an undesirable event to full system recovery. System dynamics simulation facilitates the modeling of dynamic interactions among complex arrangements of system components, providing outputs of system performance that can be used to quantify safety. This paper presents the framework for a modeling approach that can be used to simulate a range of potential operating conditions for a hydropower infrastructure system. Details of the generic hydropower infrastructure system simulation model are provided. A case study is used to evaluate system outcomes in response to a particular earthquake scenario, with two system safety performance measures shown. Results indicate that the simulation model is able to estimate potential measures of system safety which relate to flow conveyance and flow retention. A comparison of operational and upgrade strategies is shown to demonstrate the utility of the model for comparing various operational response strategies, capital upgrade alternatives, and maintenance regimes. Results show that seismic upgrades to the spillway gates provide the largest improvement in system performance for the system and scenario of interest.

  10. 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.

  11. 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.

  12. 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.

  13. Radiobiological basis for setting neutron radiation safety standards

    International Nuclear Information System (INIS)

    Straume, T.

    1985-01-01

    Present neutron standards, adopted more than 20 yr ago from a weak radiobiological data base, have been in doubt for a number of years and are currently under challenge. Moreover, recent dosimetric re-evaluations indicate that Hiroshima neutron doses may have been much lower than previously thought, suggesting that direct data for neutron-induced cancer in humans may in fact not be available. These recent developments make it urgent to determine the extent to which neutron cancer risk in man can be estimated from data that are available. Two approaches are proposed here that are anchored in particularly robust epidemiological and experimental data and appear most likely to provide reliable estimates of neutron cancer risk in man. The first approach uses gamma-ray dose-response relationships for human carcinogenesis, available from Nagasaki (Hiroshima data are also considered), together with highly characterized neutron and gamma-ray data for human cytogenetics. When tested against relevant experimental data, this approach either adequately predicts or somewhat overestimates neutron tumorigenesis (and mutagenesis) in animals. The second approach also uses the Nagasaki gamma-ray cancer data, but together with neutron RBEs from animal tumorigenesis studies. Both approaches give similar results and provide a basis for setting neutron radiation safety standards. They appear to be an improvement over previous approaches, including those that rely on highly uncertain maximum neutron RBEs and unnecessary extrapolations of gamma-ray data to very low doses. Results suggest that, at the presently accepted neutron dose limit of 0.5 rad/yr, the cancer mortality risk to radiation workers is not very different from accidental mortality risks to workers in various nonradiation occupations

  14. Westinghouse independent safety review of Savannah River production reactors

    International Nuclear Information System (INIS)

    Leggett, W.D.; McShane, W.J.; Liparulo, N.J.; McAdoo, J.D.; Strawbridge, L.E.; Call, D.W.

    1989-01-01

    Westinghouse Electric Corporation has performed a safety assessment of the Savannah River production reactors (K, L, and P) as requested by the US Department of Energy. This assessment was performed between November 1, 1988, and April 1, 1989, under the transition contract for the Westinghouse Savannah River Company's preparations to succeed E.I. du Pont de Nemours ampersand Company as the US Department of Energy contractor for the Savannah River Project. The reviewers were drawn from several Westinghouse nuclear energy organizations, embody a combination of commercial and government reactor experience, and have backgrounds covering the range of technologies relevant to assessing nuclear safety. The report presents the rationale from which the overall judgment was drawn and the basis for the committee's opinion on the phased restart strategy proposed by E.I. du Pont de Nemours ampersand Company, Westinghouse, and the US Department of Energy-Savannah River. The committee concluded that it could recommend restart of one reactor at partial power upon completion of a list of recommended upgrades both to systems and their supporting analyses and after demonstration that the organization had assimilated the massive changes it will have undergone. 37 refs., 1 fig., 3 tabs

  15. Westinghouse independent safety review of Savannah River production reactors

    Energy Technology Data Exchange (ETDEWEB)

    Leggett, W.D.; McShane, W.J. (Westinghouse Hanford Co., Richland, WA (USA)); Liparulo, N.J.; McAdoo, J.D.; Strawbridge, L.E. (Westinghouse Electric Corp., Pittsburgh, PA (USA). Nuclear and Advanced Technology Div.); Toto, G. (Westinghouse Electric Corp., Pittsburgh, PA (USA). Nuclear Services Div.); Fauske, H.K. (Fauske and Associates, Inc., Burr Ridge, IL (USA)); Call, D.W. (Westinghouse Savannah R

    1989-04-01

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

  16. Analysis of Electrical Safety Conditions Taking into Account Soil Conductivity Determined on the Basis of Fuzzy Logic

    OpenAIRE

    Manusov, V.Z.; Zaytseva, N.M.

    2017-01-01

    The goal of this work is to prove a possibility of determining soil parameters that influence its conductivity being the basis of grounding, step voltage and touch voltage calculation. This in its turn increases the safety level of electric equipment operation. The article is devoted to development of new, no conventional models of soil conductivity using the theory of fuzzy sets and fuzzy logic. The description of the solution includes the following sections: fuzzy models of specific electri...

  17. Nuclear safety activities in SR Slovenia in 1985

    International Nuclear Information System (INIS)

    1986-09-01

    Currently Yugoslavia has one 632 MWe nuclear power plant of PWR design, located at Krsko in the Socialist Republic of Slovenia. NPP Krsko, which is a two-loop plant, started power operation in 1981. In general, reactor safety activities in SR Slovenia are mostly related to upgrading the safety of our NPP Krsko and to develop capabilities to be used for the future units. This report presents safety related organizations in SR Slovenia and their activities performed in 1985. (author)

  18. Nuclear safety activities in SR Slovenia in 1985

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1986-09-15

    Currently Yugoslavia has one 632 MWe nuclear power plant of PWR design, located at Krsko in the Socialist Republic of Slovenia. NPP Krsko, which is a two-loop plant, started power operation in 1981. In general, reactor safety activities in SR Slovenia are mostly related to upgrading the safety of our NPP Krsko and to develop capabilities to be used for the future units. This report presents safety related organizations in SR Slovenia and their activities performed in 1985. (author)

  19. Implementation of an Enhanced Measurement Control Program for handling nuclear safety samples at WSRC

    International Nuclear Information System (INIS)

    Boler-Melton, C.; Holland, M.K.

    1991-01-01

    In the separation and purification of nuclear material, nuclear criticality safety (NCS) is of primary concern. The primary nuclear criticality safety controls utilized by the Savannah River Site (SRS) Separations Facilities involve administrative and process equipment controls. Additional assurance of NCS is obtained by identifying key process hold points where sampling is used to independently verify the effectiveness of production control. Nuclear safety measurements of samples from these key process locations provide a high degree of assurance that processing conditions are within administrative and procedural nuclear safety controls. An enhanced procedure management system aimed at making improvements in the quality, safety, and conduct of operation was implemented for Nuclear Safety Sample (NSS) receipt, analysis, and reporting. All procedures with nuclear safety implications were reviewed for accuracy and adequate detail to perform the analytical measurements safely, efficiently, and with the utmost quality. Laboratory personnel worked in a ''Deliberate Operating'' mode (a systematic process requiring continuous expert oversight during all phases of training, testing, and implementation) to initiate the upgrades. Thus, the effort to revise and review nuclear safety sample procedures involved a team comprised of a supervisor, chemist, and two technicians for each procedure. Each NSS procedure was upgraded to a ''Use Every Time'' (UET) procedure with sign-off steps to ensure compliance with each step for every nuclear safety sample analyzed. The upgrade program met and exceeded both the long and short term customer needs by improving measurement reliability, providing objective evidence of rigid adherence to program principles and requirements, and enhancing the system for independent verification of representative sampling from designated NCS points

  20. Physical protection upgrades in Ukraine

    International Nuclear Information System (INIS)

    Djakov, A.

    1998-01-01

    The U.S. DOE is providing nuclear material safeguards assistance in both material control and accountability and in physical protection to several facilities in Ukraine. This paper summarizes the types of physical protection upgrades that have been or are presently being implemented at these facilities. These facilities include the Kiev Institute for Nuclear Research, Kharkov Institute of Physics and Technology, Sevastopol Institute of Nuclear Energy and Industry, and the South Ukraine Nuclear Power Plant. Typical upgrades include: hardening of storage areas; improvements in access control, intrusion detection, and CCTV assessment; central alarm station improvements; and implementation of new voice communication systems. Methods used to implement these upgrades and problems encountered are discussed. Training issues are also discussed

  1. Fire safety assessment of tunnel structures

    DEFF Research Database (Denmark)

    Gkoumas, Konstantinos; Giuliani, Luisa; Petrini, Francesco

    2011-01-01

    .g. structural and non structural, organizational, human behavior). This is even more truth for the fire safety design of such structures. Fire safety in tunnels is challenging because of the particular environment, bearing in mind also that a fire can occur in different phases of the tunnel’s lifecycle. Plans...... for upgrading fire safety provisions and tunnel management are also important for existing tunnels. In this study, following a brief introduction of issues regarding the above mentioned aspects, the structural performance of a steel rib for a tunnel infrastructure subject to fire is assessed by means...

  2. Block 4 and Cluster D - Safety

    CERN Multimedia

    CERN. Geneva

    2016-01-01

    After transfer of Bloc4 test facility to SM18 its safety system was upgraded in 2014/2015 for its 3 existing test cryostats. In 2016 a new test facility HFM (High Field Magnet) was integrated in the Bloc4. A new test facility Cluster D which is similar to HFM is in a process of installation in SM18. The Presentation is dedicated to safety strategy of Bloc4 and Cluster D and focused on an analysis of risk conditions and on the real behaviour of safety valves in case of discharge. Protection of sub-atmospheric circuits is also discussed.

  3. OSR encapsulation basis -- 100-KW

    International Nuclear Information System (INIS)

    Meichle, R.H.

    1995-01-01

    The purpose of this report is to provide the basis for a change in the Operations Safety Requirement (OSR) encapsulated fuel storage requirements in the 105 KW fuel storage basin which will permit the handling and storing of encapsulated fuel in canisters which no longer have a water-free space in the top of the canister. The scope of this report is limited to providing the change from the perspective of the safety envelope (bases) of the Safety Analysis Report (SAR) and Operations Safety Requirements (OSR). It does not change the encapsulation process itself

  4. Approach to developing a ground-motion design basis for facilities important to safety at Yucca Mountain

    International Nuclear Information System (INIS)

    King, J.L.

    1990-01-01

    This paper discusses a methodology for developing a ground-motion design basis for prospective facilities at Yucca Mountain that are important to safety. The methodology utilizes a guasi-deterministic construct called the 10,000-year cumulative-slip earthquake that is designed to provide a conservative, robust, and reproducible estimate of ground motion that has a one-in-ten chance of occurring during the preclosure period. This estimate is intended to define a ground-motion level for which the seismic design would ensure minimal disruption to operations engineering analyses to ensure safe performance are included

  5. EPRI training to support digital upgrades in nuclear plants

    International Nuclear Information System (INIS)

    Torok, R.; Naser, J.

    2006-01-01

    In response to the growing challenges of obsolescence and rising maintenance costs, utilities are increasingly upgrading or replacing their existing instrumentation and control (I and C) systems and components. In most cases, this involves analog-to-digital or digital-to-newer-digital replacements. However, the use of digital technology often raises new technical and licensing issues, particularly for safety-related applications. Examples include: new failure modes and the potential for common-mode failure of redundant components; electromagnetic compatibility (EMC); potential human-system interface problems; and software verification, validation, and configuration management. Successful implementation, operation and maintenance of digital systems depend to a great extent on having processes in place that are tailored for digital technology. Nuclear plants are therefore updating their processes as they start the migration to digital I and C. For several years, EPRI has been developing guidelines to address the key technical issues and interfacing with the U.S. NRC to ensure the acceptability of the approaches developed. A framework for implementing digital upgrades now exists, but practical experience under the current regulatory environment is sparse. Significant uncertainty exists in regard to both technical and licensing issues. Many utilities still need to prepare their staff and processes to properly handle the new technology. In recent years, this need has been exacerbated by staff reductions, changing job assignments, and declining training budgets. EPRI has responded by developing a training program to help utilities efficiently bring design and licensing engineers up to speed on the latest issues and guidance affecting the implementation of digital upgrades in nuclear plants. This paper describes the key technical issues in the context of the EPRI training program. (authors)

  6. Seismic qualification of non-safety class equipment whose failure would damage safety class equipment

    International Nuclear Information System (INIS)

    LaSalle, F.R.

    1991-01-01

    Both Code of Federal Regulations, Title 10, Part 50, and US Department of Energy Order 6340.1A have requirements to assess the interaction of non-safety and safety class structures and equipment during a seismic event to maintain the safety function. At the Hanford Site, a cost effective program has been developed to perform the evaluation of non-safety class equipment. Seismic qualification is performed by analysis, test, or upgrading of the equipment to ensure the integrity of safety class structures and equipment. This paper gives a brief overview and synopsis that address design analysis guidelines including applied loading, damping values, component anchorage, allowable loads, and stresses. Test qualification of equipment and walkdown acceptance criteria for heating ampersand ventilation (H ampersand V) ducting, conduit, cable tray, missile zone of influence, as well as energy criteria are presented

  7. 29 CFR 1975.2 - Basis of authority.

    Science.gov (United States)

    2010-07-01

    ... Relating to Labor (Continued) OCCUPATIONAL SAFETY AND HEALTH ADMINISTRATION, DEPARTMENT OF LABOR (CONTINUED) COVERAGE OF EMPLOYERS UNDER THE WILLIAMS-STEIGER OCCUPATIONAL SAFETY AND HEALTH ACT OF 1970 § 1975.2 Basis... Occupational Safety and Health Act of 1970, is derived mainly from the Commerce Clause of the Constitution...

  8. Upgrading Probability via Fractions of Events

    Directory of Open Access Journals (Sweden)

    Frič Roman

    2016-08-01

    Full Text Available The influence of “Grundbegriffe” by A. N. Kolmogorov (published in 1933 on education in the area of probability and its impact on research in stochastics cannot be overestimated. We would like to point out three aspects of the classical probability theory “calling for” an upgrade: (i classical random events are black-and-white (Boolean; (ii classical random variables do not model quantum phenomena; (iii basic maps (probability measures and observables { dual maps to random variables have very different “mathematical nature”. Accordingly, we propose an upgraded probability theory based on Łukasiewicz operations (multivalued logic on events, elementary category theory, and covering the classical probability theory as a special case. The upgrade can be compared to replacing calculations with integers by calculations with rational (and real numbers. Namely, to avoid the three objections, we embed the classical (Boolean random events (represented by the f0; 1g-valued indicator functions of sets into upgraded random events (represented by measurable {0; 1}-valued functions, the minimal domain of probability containing “fractions” of classical random events, and we upgrade the notions of probability measure and random variable.

  9. Project scenarios for bitumen upgrading

    International Nuclear Information System (INIS)

    Koppel, P.E.; Mazurek, W.L.; Harji, A.

    2002-01-01

    The established reserves of Alberta's heavy oil resources are 178 billion barrels, and potential recoverable reserves are 315 billion barrels. The challenge of production includes the logistics of recovery, upgrading and transportation to market. Utilization of the bitumen is not simple because bitumen is too viscous to transport by pipeline. In addition, it is not processable by most existing refineries unless it can be upgraded through dilution. This paper examined different factors regarding the economic viability of various upgrading methods of a wide range of bitumen feedstocks. The study also examined the sensitivity of refinery demand to the prices of these feedstocks, along with the competitiveness among bitumen-based feedstock and conventional crudes. Western Canada, Ontario and the PADD II district in the United States are the 3 major markets for western Canadian bitumen based feedstock, the demand for which depends on refinery configurations and asphalt demand. This paper described the following 4 generic scenarios that describe Alberta bitumen upgrading projects: (1) adjacent to open pit mines, (2) adjacent to steam assisted gravity drainage (SAGD) facilities, (3) remotely located from resource production at an existing refinery, and (4) pipeline bitumen. It was noted that producers should determine the best way to upgrade the bitumen to ensure there is an economic market for the product, but they should also be aware not to over process the bitumen so as not to leave existing refinery facilities under-utilized. 2 refs., 1 tab., 3 figs

  10. System and process for upgrading hydrocarbons

    Science.gov (United States)

    Bingham, Dennis N.; Klingler, Kerry M.; Smith, Joseph D.; Turner, Terry D.; Wilding, Bruce M.

    2015-08-25

    In one embodiment, a system for upgrading a hydrocarbon material may include a black wax upgrade subsystem and a molten salt gasification (MSG) subsystem. The black wax upgrade subsystem and the MSG subsystem may be located within a common pressure boundary, such as within a pressure vessel. Gaseous materials produced by the MSG subsystem may be used in the process carried out within the black wax upgrade subsystem. For example, hydrogen may pass through a gaseous transfer interface to interact with black wax feed material to hydrogenate such material during a cracking process. In one embodiment, the gaseous transfer interface may include one or more openings in a tube or conduit which is carrying the black wax material. A pressure differential may control the flow of hydrogen within the tube or conduit. Related methods are also disclosed.

  11. CANDU Digital Control Computer upgrade options

    International Nuclear Information System (INIS)

    De Jong, M.S.; De Grosbois, J.; Qian, T.

    1997-01-01

    This paper reviews the evolution of Digital Control Computers (DCC) in CANDU power plants to the present day. Much of this evolution has been to meeting changing control or display requirements as well as the replacement of obsolete, or old and less reliable technology with better equipment that is now available. The current work at AECL and Canadian utilities to investigate DCC upgrade options, alternatives, and strategies are examined. The dependence of a particular upgrade strategy on the overall plant refurbishment plans are also discussed. Presently, the upgrade options range from replacement of individual obsolete system components, to replacement of the entire DCC hardware without changing the software, to complete replacement of the DCCs with a functionally equivalent system using new control computer equipment and software. Key issues, constraints and objectives associated with these DCC upgrade options are highlighted. (author)

  12. Safety Review related to Commercial Grade Digital Equipment in Safety System

    International Nuclear Information System (INIS)

    Yu, Yeongjin; Park, Hyunshin; Yu, Yeongjin; Lee, Jaeheung

    2013-01-01

    The upgrades or replacement of I and C systems on safety system typically involve digital equipment developed in accordance with non-nuclear standards. However, the use of commercial grade digital equipment could include the vulnerability for software common-mode failure, electromagnetic interference and unanticipated problems. Although guidelines and standards for dedication methods of commercial grade digital equipment are provided, there are some difficulties to apply the methods to commercial grade digital equipment for safety system. This paper focuses on regulatory guidelines and relevant documents for commercial grade digital equipment and presents safety review experiences related to commercial grade digital equipment in safety system. This paper focuses on KINS regulatory guides and relevant documents for dedication of commercial grade digital equipment and presents safety review experiences related to commercial grade digital equipment in safety system. Dedication including critical characteristics is required to use the commercial grade digital equipment on safety system in accordance with KEPIC ENB 6370 and EPRI TR-106439. The dedication process should be controlled in a configuration management process. Appropriate methods, criteria and evaluation result should be provided to verify acceptability of the commercial digital equipment used for safety function

  13. A risk-informed framework for establishing a beyond design basis safety basis for external hazards

    Energy Technology Data Exchange (ETDEWEB)

    Amico, P. [Hughes Associates, Inc, Baltimore, MD (United States); Anoba, R. [Hughes Associates, Inc, Raleigh, NC (United States); Najafi, B. [Hughes Associates, Inc., Los Gatos, CA (United States)

    2014-07-01

    The events at Fukushima Daiichi taught us that meeting a deterministic design basis requirement for external hazards does not assure that the risk is low. As observed at the plant, the two primary reasons for this are failure cliffs above the design basis event and that combined hazard effects are not considered in design. Because the possible combinations of design basis exceedences and external hazard combinations are very large and complex, an approach focusing only on the most important ones is needed. For this reason, a risk informed approach is the most effective approach, which is discussed in this paper. (author)

  14. Upgrade of the ALICE muon trigger electronics

    International Nuclear Information System (INIS)

    Dupieux, P; Joly, B; Jouve, F; Manen, S; Vandaële, R

    2014-01-01

    The ALICE muon trigger is a large scale detector based on single gap bakelite RPCs. An upgrade of the electronics is needed in order to withstand the increase of luminosity after the LHC Long Shutdown-2 in 2018-2019. The detector will be read out at the minimum bias rate of 100 kHz in Pb–Pb collisions (including a safety factor of 2), two orders of magnitude above the present design. For the most exposed RPCs and in the present conditions of operation, the total integrated charge could be as high as 100 mC/cm 2 with rates up to 100 Hz/cm 2 , which is above the present limit for safe operation. In order to overcome these limitations, upgrade projects of the Front-End (FE) and Readout Electronics are scheduled. The readout upgrade at high rate with low dead time requires changing most of the present electronics. It involves a new design for the 234 Local cards receiving the LVDS signals from the FE electronics and the 16 Regional concentrator cards. The readout chain is completed by a single Common Readout Unit developed for most ALICE sub-detectors. The new architecture of the muon trigger readout will be briefly presented. The present FE electronics, designed for the streamer mode, must be replaced to prevent ageing of the RPCs in the future operating conditions. The new FE called FEERIC (for Front-End Electronics Rapid Integrated Circuit) will have to perform amplification of the analog input signals. This will allow for RPC operation in a low-gain avalanche mode, with a much smaller charge deposit (factor 3-5) in the detector as compared to the present conditions. The purpose is to discriminate RPC signals with a charge threshold around 100 fC, in both polarities, and with a time jitter below 1 ns. We will describe the FE card and FEERIC ASIC features and first prototype performance, report on test results obtained on a cosmic test bench and discuss ongoing developments

  15. PSI: Upgrading

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    The accelerator complex at the Paul Scherrer Institute in Villigen near Zurich (PSI - formed in 1988 by combining the Federal Institute for Reactor Research and the Swiss Institute for Nuclear Research) is in the throes of a major and lengthy upgrade

  16. PSI: Upgrading

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1992-09-15

    The accelerator complex at the Paul Scherrer Institute in Villigen near Zurich (PSI - formed in 1988 by combining the Federal Institute for Reactor Research and the Swiss Institute for Nuclear Research) is in the throes of a major and lengthy upgrade.

  17. Operational test report for the 241-A-701 air compressor upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Meeuwsen, W.E.

    1997-06-30

    A description and safety class designation of the accumulator and 701-A compressor system is contained in VTHC-SD-@-DA-137, Safety Classification (of the 241-A-70) Compressed Air System and shown on drawings H-2-62895, Sheet 2 and H-14-20308, Sheet 3. The design basis for the 241-A-702 Ventilation System Accumulator is contained in @-C-SD-@-DB-016, 241-A-702 Ventilation System Accumulator Design Basis.

  18. The Argentine Approach to Radiation Safety: Its Ethical Basis

    International Nuclear Information System (INIS)

    Gonzalez, A.J.

    2011-01-01

    The ethical bases of Argentina's radiation safety approach are reviewed. The applied principles are those recommended and established internationally, namely: the principle of justification of decisions that alters the radiation exposure situation; the principle of optimization of protection and safety; the principle of individual protection for restricting possible inequitable outcomes of optimized safety; and the implicit principle of inter generational prudence for protection future generations and the habitat. The principles are compared vis-a-vis the prevalent ethical doctrines: justification vis-a-vis teleology; optimization vis-a-vis utilitarianism; individual protection vis-a-vis de ontology; and, inter generational prudence vis-a-vis aretaicism (or virtuosity). The application of the principles and their ethics in Argentina is analysed. These principles are applied to All exposure to radiation harm; namely, to exposures to actual doses and to exposures to actual risk and potential doses, including those related to the safety of nuclear installations, and they are harmonized and applied in conjunction. It is concluded that building a bridge among all available ethical doctrines and applying it to radiation safety against actual doses and actual risk and potential doses is at the roots of the successful nuclear regulatory experience in Argentina.

  19. Biogas upgrading to biomethane. Proceedings; Biogasaufbereitung zu Biomethan. Tagungsband

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2008-12-05

    Within the 6th Hanauer Dialogue 'Biogas upgrading to biomethane' at 21st February, 2008, the following lectures were held: (a) Processing of biogas - an introduction (Michael Beil); (b) The climate protecting targets of the Federal Republic of Germany: Which role will play the upgrading of biogas, and which legal boundary conditions are created by the Federal Government? (Uwe Holzhammer); (c) Future strategy: CH{sub 4} grids (Juergen Schmid); (d) Biogas upgrading and biomethane utilization in Sweden (Anneli Petersson); (e) Biogas upgrading and utilization of bio methane in Switzerland (Arthur Wellinger); (f) Biogas upgrading by means of pressure swing adsorption (Alfons Schulte-Schulze Berndt); (g) Biogas upgrading by means of pressurized water washing (Ulf Richter); (h) Biogas upgrading for feeding in public grids. The case of biogas plant Bruck a.d. Leitha (Michael Harasek); (i) Biogas upgrading by means of chemical absorption according to the LP Cooab process (Jerome van Beek); (j) Practical experiences in unpressurized amine washing MT bio methane (Karsten Wuensche); (k) Biogas upgrading by means of organic physical washing with HAASE biogas amplifiers (Roland Kahn); (l) Upgrading using cryogenic technology; the GPP registered -system (Jeroen de Pater); (m) Micro Gas Distribution Systems: Alternatives to biogas upgrading and grid injection (Michael Beil, Bernd Krautkremer); (n) Feeding of exchange gas. The case of project Straelen and Kerpen (Frank Schaefer); (o) Feeding of biogas from the view of grid operators (Norbert Nordmeyer); BIOGASMAX: Biogas as Vehicle Fuel - Market Expansion to 2020 Air Quality (Michael Beil, Uwe Hoffstede); (p) Study: Feeding of biogas into the natural gas distribution system (Fachagentur Nachwachsende Rohstoffe).

  20. Seismic analysis of two 1050 mm diameter heavy water upgrading towers for 235 MWe Kaiga Atomic Power Plant Site

    International Nuclear Information System (INIS)

    Soni, R.S.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.; Narwaria, Suresh; Vardarajan, T.G.; Sadhukhan, H.K.

    1992-01-01

    This report deals with the analysis carried out for the evaluation of earthquake induced stresses and deflections in two 1050 mm diameter heavy water upgrading towers for Kaiga Atomic Power Plant Site. The analysis of upgrading tower has been carried out for two mutually perpendicular horizontal excitations and one vertical excitation applied simultaneously. The upgrading towers have been analysed using beam model taking into account soil-structure interaction. Response spectrum analysis has been carried out using site spectra for 235 MWe Kaiga reactor site. The seismic analysis has been performed for both the towers with supporting structure along with concrete pedestals and raft foundation. The towers have been checked for its stability due to compressive stresses to avoid buckling so that the nearby safety related structures are not geopardised in the event of safe shutdown earthquake (SSE) loading. (author). 14 refs., 12 figs., 18 tabs

  1. The D0 detector upgrade

    International Nuclear Information System (INIS)

    Bross, A.D.

    1995-02-01

    The Fermilab collider program is undergoing a major upgrade of both the accelerator complex and the two detectors. Operation of the Tevatron at luminosities upwards of ten time that currently provided will occur in early 1999 after the commissioning of the new Fermilab Main Injector. The D0 upgrade program has been established to deliver a detector that will meet the challenges of this environment. A new magnetic tracker consisting of a superconducting solenoid, a silicon vertex detector, a scintillating fiber central tracker, and a central preshower detector will replace the current central tracking and transition radiation chambers. We present the design and performance capabilities of these new systems and describe results from physics simulations that demonstrate the physics reach of the upgraded detector

  2. Public-Interest Benefit Evaluation of Partial- Upgrading Technology

    Directory of Open Access Journals (Sweden)

    G. Kent Fellows

    2017-01-01

    Full Text Available Approximately 60 per cent of Alberta’s oil sands production is non-upgraded bitumen which, after being mixed with a diluting agent (diluent to allow transport, is exported. A popular view within Alberta — and particularly among Albertan politicians — is that a much larger share of oil sands bitumen should be upgraded in the province. However, without public subsidies or government underwriting, it is uneconomic to build and operate new facilities in Alberta to fully upgrade the bitumen into synthetic crude oil. But there are new partial upgrading technologies being developed that, subject to successful testing at a larger (commercial pilot scale, can prove to be not only economic in Alberta, but also generate large social and economic benefits for the province. The advantages include a much smaller capital investment, a significant increase in the value of the product and market for the product and, even more importantly, a dramatic reduction in the need for large amounts of expensive diluent to transport the product to market. Indeed, the only diluent required will be that to move the bitumen from the production site to the partial upgrader and this can be continually recycled. The market for the synthetic crude oil produced by full upgrading is only getting tougher. Any Alberta bitumen fully upgraded here would compete closely with the rapidly expanding supply of light U.S. unconventional oil. Partial upgrading does not upgrade bitumen to a light crude, but to something resembling more of a medium or heavy crude, and at a lower cost per barrel than full upgrading. Unlike in the increasingly crowded light-crude market, the Alberta Royalty Review Advisory Panel recognized that currently there are gaps in several North American refineries that could be filled by this partially upgraded Alberta oil. A partial upgrader serving that less-competitive market not only appears to hold the potential for investors to make attractive returns in the

  3. Operating experience and systems analysis at Trillo NPP: A program intended for systematic review of plant safety systems to assess design basis requirements compliance

    International Nuclear Information System (INIS)

    Vega, R. de la

    1996-01-01

    The program was defined to apply to all plant safety systems and/or systems included in plant Technical Specifications. The goal of the program was to ensure, by systematic design, construction, and commissioning review, the adequacy of safety systems, structures and components to fulfill their safety functions. Also, as a result of the program, it was established that a complete, unambiguous, systematic, design basis definition shall take place. And finally, a complete documental review of the plant design shall result from the program execution

  4. TCV divertor upgrade for alternative magnetic configurations

    Directory of Open Access Journals (Sweden)

    H. Reimerdes

    2017-08-01

    Full Text Available The Swiss Plasma Center (SPC is planning a divertor upgrade for the TCV tokamak. The upgrade aims at extending the research of conventional and alternative divertor configurations to operational scenarios and divertor regimes of greater relevance for a fusion reactor. The main elements of the upgrade are the installation of an in-vessel structure to form a divertor chamber of variable closure and enhanced diagnostic capabilities, an increase of the pumping capability of the divertor chamber and the addition of new divertor poloidal field coils. The project follows a staged approach and is carried out in parallel with an upgrade of the TCV heating system. First calculations using the EMC3-Eirene code indicate that realistic baffles together with the planned heating upgrade will allow for a significantly higher compression of neutral particles in the divertor, which is a prerequisite to test the power dissipation potential of various divertor configurations.

  5. Initial performance of upgraded Tevatron cryogenic systems

    International Nuclear Information System (INIS)

    Norris, B.L.

    1996-09-01

    Fermilab began operating a re-designed satellite refrigerator systems in November 1993. Upgrades were installed to operate the Tevatron at a magnet temperature of 3.5 K, approximately 1K lower than the original design. Refrigerator upgrades included new valve boxes, larger reciprocating expanders, the installation of cold vapor compressors, new sub-atmospheric instrumentation and an entirely new distributed controls system. Cryogenic system reliability data for Colliding Physics Run 1B is presented emphasizing a failure analysis for each aspect of the upgrade. Comparison to data for Colliding Physics Run 1A (previous to upgrade) is presented to show the impact of a major system overhaul. New operational problems and their solutions are presented in detail

  6. Operational test report for the 241-A-701 air compressor upgrade

    International Nuclear Information System (INIS)

    Meeuwsen, W.E.

    1997-01-01

    A description and safety class designation of the accumulator and 701-A compressor system is contained in VTHC-SD-at sign-DA-137, Safety Classification (of the 241-A-70) Compressed Air System and shown on drawings H-2-62895, Sheet 2 and H-14-20308, Sheet 3. The design basis for the 241-A-702 Ventilation System Accumulator is contained in at sign-C-SD-at sign-DB-016, 241-A-702 Ventilation System Accumulator Design Basis

  7. Analysis of Energy Industry Upgrading in Northeast China

    Science.gov (United States)

    Liu, Xiao-jing; Ji, Yu-liang; Guan, Bai-feng; Jing, Xin

    2018-02-01

    Promoting regional economic growth and realizing the transformation of the mode of economic growth are in industrial upgrading essence The product is a carrier that represents a series of links of production, management and marketing behind the enterprise, and is a comprehensive reflection of the knowledge and ability of a country or region. Based on the industrial spatial structure, this paper visualizes the industrial space in Northeast China from 2005 to 2015, analyzes the comparative advantages of the energy industry in Northeast China, and examines the status quo of the upgrade of the energy industry according to the industrial upgrading status. Based on the industrial spatial structure, Industry intensity in the industrial space, put forward the future direction of the energy industry upgrade and upgrade path.

  8. Developments towards the LHCb VELO upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Cid Vidal, Xabier, E-mail: xabier.cid.vidal@cern.ch

    2016-09-21

    The Vertex Locator (VELO) is a silicon strip detector surrounding the interaction region of the LHCb experiment. The upgrade of the VELO is planned to be installed in 2019–2020, and the current detector will be replaced by a hybrid pixel system equipped with electronics capable of reading out at a rate of 40 MHz. The new detector is designed to withstand the radiation dose expected at an integrated luminosity of 50 fb{sup −1}. The detector will be composed of silicon pixel sensors, read out by the VeloPix ASIC that is being developed based on the TimePix/MediPix family. The prototype sensors for the VELO upgrade are being irradiated in five different facilities and the post-irradiation performance is being measured with testbeams, and in the lab. These proceedings present the VELO upgrade and briefly discuss the results of the sensor testing campaign. - Highlights: • The VELO is the detector surrounding the LHCb collision point. • After its upgrade, the VELO will be capable of reading out at a rate of 40 MHz. • The detector will be composed of Si pixel sensors, read out by the VeloPix ASIC. • The irradiated prototype sensors for the VELO upgrade are currently being tested.

  9. Alternative off-site power supply improves nuclear power plant safety

    International Nuclear Information System (INIS)

    Gjorgiev, Blaže; Volkanovski, Andrija; Kančev, Duško; Čepin, Marko

    2014-01-01

    Highlights: • Additional power supply for mitigation of the station blackout event in NPP is used. • A hydro power plant is considered as an off-site alternative power supply. • An upgrade of the probabilistic safety assessment from its traditional use is made. • The obtained results show improvement of nuclear power plant safety. - Abstract: A reliable power system is important for safe operation of the nuclear power plants. The station blackout event is of great importance for nuclear power plant safety. This event is caused by the loss of all alternating current power supply to the safety and non-safety buses of the nuclear power plant. In this study an independent electrical connection between a pumped-storage hydro power plant and a nuclear power plant is assumed as a standpoint for safety and reliability analysis. The pumped-storage hydro power plant is considered as an alternative power supply. The connection with conventional accumulation type of hydro power plant is analysed in addition. The objective of this paper is to investigate the improvement of nuclear power plant safety resulting from the consideration of the alternative power supplies. The safety of the nuclear power plant is analysed through the core damage frequency, a risk measure assess by the probabilistic safety assessment. The presented method upgrades the probabilistic safety assessment from its common traditional use in sense that it considers non-plant sited systems. The obtained results show significant decrease of the core damage frequency, indicating improvement of nuclear safety if hydro power plant is introduced as an alternative off-site power source

  10. RF upgrade program in LHC injectors and LHC machine

    International Nuclear Information System (INIS)

    Jensen, E.

    2012-01-01

    The main themes of the RF upgrade program are: the Linac4 project, the LLRF-upgrade and the study of a tuning-free wide-band system for PSB, the upgrade of the SPS 800 MHz amplifiers and beam controls and the upgrade of the transverse dampers of the LHC. Whilst LHC Splice Consolidation is certainly the top priority for LS1, some necessary RF consolidation and upgrade is necessary to assure the LHC performance for the next 3- year run period. This includes: 1) necessary maintenance and consolidation work that could not fit the shorter technical stops during the last years, 2) the upgrade of the SPS 200 MHz system from presently 4 to 6 cavities and possibly 3) the replacement of one LHC cavity module. On the longer term, the LHC luminosity upgrade requires crab cavities, for which some preparatory work in SPS Coldex must be scheduled during LS1. (author)

  11. OMEGA upgrade staging options

    International Nuclear Information System (INIS)

    Kelly, J.H.; Shoup, M.J.; Smith, D.L.

    1989-01-01

    The authors discuss how they are designing an upgrade to its 24-beam OMEGA laser system, OMEGA is a frequency tripled, all-rod system capable of producing 2 kJ at 0.8 ns on target. Important direct-drive-target-ignition physics could be investigated with an upgraded system capable of producing a shaped pulse consisting of a long (5ns) low-intensity, foot, smoothly transitioning into a short (0.5 ns), intense, compression pulse. The total pulse energy is 30 kJ, which, from target-irradiation uniformity considerations, must be distributed over 60 beams

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

    International Nuclear Information System (INIS)

    Cook, David Howard

    2009-01-01

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

  13. Upgrading the ATLAS Tile Calorimeter electronics

    CERN Document Server

    Oreglia, M; The ATLAS collaboration

    2013-01-01

    The Tile Calorimeter (TileCal) is the hadronic calorimeter covering the most central region of the ATLAS experiment at LHC. The TileCal readout consists of about 10000 channels. The main upgrade will occur for the High Luminosity LHC phase (phase 2) which is scheduled around 2022. The upgrade aims at replacing the majority of the on- and off- detector electronics so that all calorimeter signals are directly digitized and sent to the off-detector electronics in the counting room. An ambitious upgrade development program is pursued studying different electronics options. Three different options are presently being investigated for the front-end electronic upgrade. Which one to use will be decided after extensive test beam studies. High speed optical links are used to read out all digitized data to the counting room. For the off-detector electronics a new back-end architecture is being developed, including the initial trigger processing and pipeline memories. A demonstrator prototype read-out for a slice of the ...

  14. Prioritizing and scheduling Portsmouth Gaseous Diffusion Plant safeguards upgrades

    International Nuclear Information System (INIS)

    Edmunds, T.; Saleh, R.; Zevanove, S.

    1992-02-01

    As part of the Site Safeguards and Security Plan (SSSP), facilities are required to develop a Resource Plan (RP). The Resource Plan provides documentation and justification for the facility's planned upgrades, including the schedule, priority, and cost estimates for the safeguards and security upgrades. Portsmouth Gaseous Diffusion Plant (PORTS) management has identified and obtained funding approval for a number of safeguards and security upgrades, including line-item construction projects. These upgrade projects were selected to address a variety of concerns identified in the PORTS vulnerability assessments and other reviews performed in support of the SSSP process. However, budgeting and scheduling constraints do not make it possible to simultaneously begin implementation of all of the upgrade projects. A formal methodology and analysis are needed to explicitly address the trade-offs between competing safeguards objectives, and to prioritize and schedule the upgrade projects to ensure that the maximum benefit can be realized in the shortest possible time frame. The purpose of this report is to describe the methodology developed to support these upgrade project scheduling decisions. The report also presents the results obtained from applying the methodology to a set of the upgrade projects selected by PORTS S ampersand S management. Data for the analysis are based on discussions with personnel familiar with the PORTS safeguards and security needs, the requirements for implementing these upgrades, and upgrade funding limitations. The analysis results presented here assume continued highly enriched uranium (HEU) operations at PORTS. However, the methodology developed is readily adaptable for the evaluation of other operational scenarios and other resource allocation issues relevant to PORTS

  15. Silicon Sensors for the Upgrades of the CMS Pixel Detector

    CERN Document Server

    Centis Vignali, Matteo; Schleper, Peter

    2015-01-01

    The Compact Muon Solenoid (CMS) is a general purpose detector at the Large Hadron Collider (LHC). The LHC luminosity is constantly increased through upgrades of the accel- erator and its injection chain. Two major upgrades will take place in the next years. The rst upgrade involves the LHC injector chain and allows the collider to achieve a luminosity of about 2 10 34 cm-2 s-1 A further upgrade of the LHC foreseen for 2025 will boost its luminosity to 5 10 34 cm-2 s1. As a consequence of the increased luminosity, the detectors need to be upgraded. In particular, the CMS pixel detector will undergo two upgrades in the next years. The rst upgrade (phase I) consists in the substitution of the current pixel detector in winter 2016/2017. The upgraded pixel detector will implement new readout elec- tronics that allow ecient data taking up to a luminosity of 2 10 34 cm-2s-1,twice as much as the LHC design luminosity. The modules that will constitute the upgraded detector are being produced at dierent institutes. Ham...

  16. Upgrading of Intermediate Bio-Oil Produced by Catalytic Pyrolysis

    Energy Technology Data Exchange (ETDEWEB)

    Abdullah, Zia [Battelle Memorial Inst., Columbus, OH (United States); Chadwell, Brad [Battelle Memorial Inst., Columbus, OH (United States); Taha, Rachid [Battelle Memorial Inst., Columbus, OH (United States); Hindin, Barry [Battelle Memorial Inst., Columbus, OH (United States); Ralston, Kevin [Battelle Memorial Inst., Columbus, OH (United States)

    2015-06-30

    The objectives of this project were to (1) develop a process to upgrade catalytic pyrolysis bio-oil, (2) investigate new upgrading catalysts suited for upgrading catalytic pyrolysis bio-oil, (3) demonstrate upgrading system operation for more than 1,000 hours using a single catalyst charge, and (4) produce a final upgraded product that can be blended to 30 percent by weight with petroleum fuels or that is compatible with existing petroleum refining operations. This project has, to the best of our knowledge, for the first time enabled a commercially viable bio-oil hydrotreatment process to produce renewable blend stock for transportation fuels.

  17. Upgrades to meet LANL SF, 121-2011, hazardous waste facility permit requirements

    International Nuclear Information System (INIS)

    French, Sean B.; Johns-Hughes, Kathryn W.

    2011-01-01

    Members of San IIdefonso have requested information from LANL regarding implementation of the revision to LANL's Hazardous Waste Facility Permit (the RCRA Permit). On January 26, 2011, LANL staff from the Waste Disposition Project and the Environmental Protection Division will provide a status update to Pueblo members at the offices of the San IIdefonso Department of Environmental and Cultural Preservation. The Waste Disposition Project presentation will focus on upgrades and improvements to LANL waste management facilities at TA-50 and TA-54. The New Mexico Environment Department issued LANL's revised Hazardous Waste Facility permit on November 30, 2010 with a 30-day implementation period. The Waste Disposition Project manages and operates four of LANL's permitted facilities; the Waste Characterization, Reduction and Repackaging Facility (WCRRF) at TA-SO, and Area G, Area L and the Radioassay and Nondestructive Testing facility (RANT) at TA-54. By implementing a combination of permanent corrective action activities and shorter-term compensatory measures, WDP was able to achieve functional compliance on December 30, 2010 with new Permit requirements at each of our facilities. One component of WOP's mission at LANL is centralized management and disposition of the Laboratory's hazardous and mixed waste. To support this mission objective, WOP has undertaken a project to upgrade our facilities and equipment to achieve fully compliant and efficient waste management operations. Upgrades to processes, equipment and facilities are being designed to provide defense-in-depth beyond the minimum, regulatory requirements where worker safety and protection of the public and the environment are concerned. Upgrades and improvements to enduring waste management facilities and operations are being designed so as not to conflict with future closure activities at Material Disposal Area G and Material Disposal Area L.

  18. Technology Solutions Case Study: Combustion Safety Simplified Test Protocol

    Energy Technology Data Exchange (ETDEWEB)

    L. Brand, D. Cautley, D. Bohac, P. Francisco, L. Shen, and S. Gloss

    2015-12-01

    Combustions safety is an important step in the process of upgrading homes for energy efficiency. There are several approaches used by field practitioners, but researchers have indicated that the test procedures in use are complex to implement and provide too many false positives. Field failures often mean that the house is not upgraded until after remediation or not at all, if not include in the program. In this report the PARR and NorthernSTAR DOE Building America Teams provide a simplified test procedure that is easier to implement and should produce fewer false positives.

  19. WNP-2 core model upgrade

    International Nuclear Information System (INIS)

    Golightly, C.E.; Ravindranath, T.K.; Belblidia, L.A.; O'Farrell, D.; Andersen, P.S.

    2006-01-01

    The paper describes the core model upgrade of the WNP-2 training simulator and the reasons for the upgrade. The core model as well as the interface with the rest of the simulator are briefly described . The paper also describes the procedure that will be used by WNP-2 to update the simulator core data after future core reloads. Results from the fully integrated simulator are presented. (author)

  20. Upgrades to the ATLAS trigger system   

    CERN Document Server

    AUTHOR|(INSPIRE)INSPIRE-00221618; The ATLAS collaboration

    2017-01-01

    In coming years the LHC is expected to undergo upgrades to increase both the energy of proton-proton collisions and the instantaneous luminosity. In order to cope with these more challenging LHC conditions, upgrades of the ATLAS trigger system will be required. This talk will focus on some of the key aspects of these upgrades. Firstly, the upgrade period between 2019-2021 will see an increase in instantaneous luminosity to $3\\times10^{34} \\rm{cm^{-2}s^{-1}}$. Upgrades to the Level 1 trigger system during this time will include improvements for both the muon and calorimeter triggers. These include the upgrade of the first-level Endcap Muon trigger, the calorimeter trigger electronics and the addition of new calorimeter feature extractor hardware, such as the Global Feature Extractor (gFEX). An overview will be given on the design and development status the aforementioned systems, along with the latest testing and validation results. By 2026, the High Luminosity LHC will be able to deliver 14 TeV collisions wit...

  1. 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

  2. The LHCb VELO Upgrade

    CERN Document Server

    de Capua, Stefano

    2016-01-01

    The upgrade of the LHCb experiment, scheduled for LHC Run-3, will transform the experiment to a triggerless system reading out the full detector at 40 MHz event rate. All data reduction algorithms will be executed in a high-level software farm, enabling the detector to run at luminosities of 2×1033 cm−2 s −1 . The Vertex Locator (VELO) is the silicon vertex detector surrounding the interaction region. The current strip detector will be replaced with a hybrid pixel system equipped with electronics capable of reading out at 40 MHz. The upgraded VELO will allow for fast pattern recognition and track reconstruction in the software trigger. The silicon pixel sensors have 55×55 µm2 pitch, and are read out by the VeloPix ASIC. The VeloPix builds on the currently available Timepix3, modified to deliver a radiation hard design capable of an order of magnitude increase in output rate. The hottest regions will have pixel hit rates of 900 Mhits/s, yielding a total data rate more than 3 Tbit/s for the upgraded VELO...

  3. The D0 upgrade

    International Nuclear Information System (INIS)

    Tuts, P.M.

    1992-10-01

    The original D0 detector was proposed in 1983, with a focus on high P T physics using precision measurements of e's, μ's, jets, and missing E T . This detector, as of the summer of 1992, has started data taking at the Fermilab Collider. However, by 1995/6 the luminosity will reach 10 31 cm -2 sec -1 , and the minimum bunch spacing will drop to 396ns from the present 3.5μs (by the Main Injector era, luminosities will approach 10 32 cm -2 sec -1 and minimum bunch spacings may reach 132ns). These changes in the accelerator conditions force us to upgrade or replace a number of detector subsystems in order to meet these new demands. In addition, the upgrade offers us the opportunity to expand the physics horizons to include not only the all important high P T physics menu, but also the low P T physics that has become increasingly important. In this paper we describe the D0 detector upgrade

  4. LHCb PID Upgrade Technical Design Report

    CERN Document Server

    LHCb Collaboration

    2013-01-01

    The LHCb upgrade will take place in the second long shutdown of the LHC, currently scheduled to begin in 2018. The upgrade will enable the experiment to run at luminosities of $2 \\times 10^{33}cm^{-2}s^{-1}$ and will read out data at a rate of 40MHz into a exible software-based trigger. All sub-detectors of LHCb will be re-designed to comply with these new operating conditions. This Technical Design Report presents the upgrade plans of the Ring Imaging Cherenkov (RICH) system, the calorimeter system and the muon system, which together provide the particle identication capabilities of the experiment.

  5. Physical safety: using of multimedia technology

    International Nuclear Information System (INIS)

    Munoz Garcia, M.

    1995-01-01

    For managing physical safety systems in nuclear power plants sophisticated computer equipment is essential. Standard safety procedure requires systems distributed throughout the plant, especially at security posts: Entrance Door, Access Control Center, Restricted Areas, etc. Besides being highly distributed, computerized physical safety management systems have always been characterized by the need to integrate very different elements. Graphic images have always played a very important role: photographs, fingerprints, signatures, etc. This makes integrated multimedia modules the ideal solution. This presentation describes how safety systems can combine not only still images, but also other multimedia data such as sound and video. It examines methodologies for using this new technology to upgrade high cost computer systems already in operation that can not be replaced. (Author)

  6. Instrumentation and control upgrade evaluation methodology: Final report. Volume 2: Workbook

    Energy Technology Data Exchange (ETDEWEB)

    Bliss, M.; Brown, E.; Florio, F.; Stofko, M.

    1996-07-01

    This workbook accompanies the methodology manual (EPRI TR-104963-V1) which describes how to develop an Upgrade Evaluation Report (UER). A UER is an evaluation that is performed by a nuclear power plant to decide the most cost-effective upgrade to perform (if any) for a previously identified Upgrade Candidate System. A UER defines the utility`s mission and objectives in regards to upgrade candidates, as well as the systems initial costs, benefits to each upgrade, and an initial upgrade schedule to cost-effectively implement system upgrades.

  7. Proceedings of the First Seminar on Radiation Safety Technology and Nuclear Biomedicine

    International Nuclear Information System (INIS)

    Suprihadi, Topo

    2003-01-01

    The First Seminar on Radiation Safety Technology and Nuclear Biomedicine was held on 10-11 April 2001 at the Center for Research and Development of Radiation Safety and Nuclear Biomedicine have presented 19 papers about upgrading manpower resources, researcher, investigator, manager, and user of nuclear facilities, to go out against free market era

  8. Climate balance of biogas upgrading systems

    International Nuclear Information System (INIS)

    Pertl, A.; Mostbauer, P.; Obersteiner, G.

    2010-01-01

    One of the numerous applications of renewable energy is represented by the use of upgraded biogas where needed by feeding into the gas grid. The aim of the present study was to identify an upgrading scenario featuring minimum overall GHG emissions. The study was based on a life-cycle approach taking into account also GHG emissions resulting from plant cultivation to the process of energy conversion. For anaerobic digestion two substrates have been taken into account: (1) agricultural resources and (2) municipal organic waste. The study provides results for four different upgrading technologies including the BABIU (Bottom Ash for Biogas Upgrading) method. As the transport of bottom ash is a critical factor implicated in the BABIU-method, different transport distances and means of conveyance (lorry, train) have been considered. Furthermore, aspects including biogas compression and energy conversion in a combined heat and power plant were assessed. GHG emissions from a conventional energy supply system (natural gas) have been estimated as reference scenario. The main findings obtained underlined how the overall reduction of GHG emissions may be rather limited, for example for an agricultural context in which PSA-scenarios emit only 10% less greenhouse gases than the reference scenario. The BABIU-method constitutes an efficient upgrading method capable of attaining a high reduction of GHG emission by sequestration of CO 2 .

  9. Disruption studies in ASDEX Upgrade

    International Nuclear Information System (INIS)

    Pautasso, G.

    2002-01-01

    Disruption generate large thermal and mechanical stresses on the tokamak components. For a future reactor disruptions have a significant impact on the design since all loading conditions must be analyzed in accordance with stricter design criteria (due to safety or difficult maintenance). Therefore the uncertainties affecting the predicted stresses must be reduced as much as possible with a more comprehensive set of measurements and analyses in this generation of experimental machines, and avoidance/ predictive methods must be developed further. The study of disruptions on ASDEX Upgrade is focused on these subjects, namely on: (1) understanding the physical mechanisms leading to this phenomenon and learning to avoid it or to predict its occurrence (with neural networks, for example) and to mitigate its effects; (2) analyzing the effects of disruptions on the machine to determine the functional dependence of the thermal and mechanical loads upon the discharge parameters. This allows to dimension or reinforce the machine components to withstand these loads and to extrapolate them to tokamaks still in the design phase; (3) learning to mitigate the consequence of disruptions. (author)

  10. Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

    Science.gov (United States)

    Emrich, William J., Jr.

    2014-01-01

    corrosion of certain potential NTR fuel forms. Additional diagnostic upgrades included in the present NTREES set up include the addition of a gamma ray spectrometer located near the vent filter to detect uranium fuel particles exiting the fuel element in the propellant exhaust stream to provide additional information any material loss occurring during testing. Other aspects of the upgrade included reworking NTREES to reduce the operational complexity of the system despite the increased complexity of the induction heating system. To this end, many of the controls were consolidated on fewer panels. As part of this upgrade activity, the Safety Assessment (SA) and the Standard Operating Procedures (SOPs) for NTREES were extensively rewritten. The new 1.2 MW induction heater consists of three physical units consisting of a transformer, rectifier, and inverter. This multiunit arrangement facilitated increasing the flexibility of the induction heater by more easily allowing variable frequency operation. Frequency ranges between 20 and 60 kHz can be accommodated in the new induction heater allowing more representative power distributions to be generated within the test elements.

  11. EDF Contribution to the Mochovce completion and safety upgrading project

    International Nuclear Information System (INIS)

    Norvez, G.

    1997-01-01

    The challenge that SE a.s. took up in April 1996 to start Mochovce units 1 and 2 within a very demanding time frame and in satisfactory conditions is on the way to being wined. EMO managed to put in place an organisation able to ensure the continuity of the design and construction works and to integrate the requirements of modern safety. EDF is proud and happy to contribute to this effort by making available its expertise and experience in building and operating nuclear power plants. The success of Mochovce will not only contribute to the Slovak electricity needs coverage, but will be the first demonstration of a soviet design nuclear power plant brought to international safety standards. (author)

  12. Design summary of the magnet support structures for the proton storage ring injection line upgrade

    International Nuclear Information System (INIS)

    Bernardin, J.D.; Ledford, J.E.; Smith, B.G.

    1997-05-01

    This report summarizes the technical engineering and design issues associated with the Proton Storage Ring (PSR) Injection Line upgrade of the Los Alamos Neutron Science Center (LANSCE). The main focus is on the engineering design calculations of several magnet support structures. The general procedure based upon a set number of design criteria is outlined, followed by a case-by-case summary of the engineering design analyses, reutilization or fabrication callouts and design safety factors

  13. Advances in boronization on NSTX-Upgrade

    Directory of Open Access Journals (Sweden)

    C. H Skinner

    2017-08-01

    Full Text Available Boronization has been effective in reducing plasma impurities and enabling access to higher density, higher confinement plasmas in many magnetic fusion devices. The National Spherical Torus eXperiment, NSTX, has recently undergone a major upgrade to NSTX-U in order to develop the physics basis for a ST-based Fusion Nuclear Science Facility (FNSF with capability for double the toroidal field, plasma current, and NBI heating power and increased pulse duration from 1–1.5s to 5–8s. A new deuterated tri-methyl boron conditioning system was implemented together with a novel surface analysis diagnostic. We report on the spatial distribution of the boron deposition versus discharge pressure, gas injection and electrode location. The oxygen concentration of the plasma facing surface was measured by in-vacuo XPS and increased both with plasma exposure and with exposure to trace residual gases. This increase correlated with the rise of oxygen emission from the plasma.

  14. Is road safety management linked to road safety performance?

    Science.gov (United States)

    Papadimitriou, Eleonora; Yannis, George

    2013-10-01

    This research aims to explore the relationship between road safety management and road safety performance at country level. For that purpose, an appropriate theoretical framework is selected, namely the 'SUNflower' pyramid, which describes road safety management systems in terms of a five-level hierarchy: (i) structure and culture, (ii) programmes and measures, (iii) 'intermediate' outcomes'--safety performance indicators (SPIs), (iv) final outcomes--fatalities and injuries, and (v) social costs. For each layer of the pyramid, a composite indicator is implemented, on the basis of data for 30 European countries. Especially as regards road safety management indicators, these are estimated on the basis of Categorical Principal Component Analysis upon the responses of a dedicated road safety management questionnaire, jointly created and dispatched by the ETSC/PIN group and the 'DaCoTA' research project. Then, quasi-Poisson models and Beta regression models are developed for linking road safety management indicators and other indicators (i.e. background characteristics, SPIs) with road safety performance. In this context, different indicators of road safety performance are explored: mortality and fatality rates, percentage reduction in fatalities over a given period, a composite indicator of road safety final outcomes, and a composite indicator of 'intermediate' outcomes (SPIs). The results of the analyses suggest that road safety management can be described on the basis of three composite indicators: "vision and strategy", "budget, evaluation and reporting", and "measurement of road user attitudes and behaviours". Moreover, no direct statistical relationship could be established between road safety management indicators and final outcomes. However, a statistical relationship was found between road safety management and 'intermediate' outcomes, which were in turn found to affect 'final' outcomes, confirming the SUNflower approach on the consecutive effect of each layer

  15. Seismic methodology in determining basis earthquake for nuclear installation

    International Nuclear Information System (INIS)

    Ameli Zamani, Sh.

    2008-01-01

    Design basis earthquake ground motions for nuclear installations should be determined to assure the design purpose of reactor safety: that reactors should be built and operated to pose no undue risk to public health and safety from earthquake and other hazards. Regarding the influence of seismic hazard to a site, large numbers of earthquake ground motions can be predicted considering possible variability among the source, path, and site parameters. However, seismic safety design using all predicted ground motions is practically impossible. In the determination of design basis earthquake ground motions it is therefore important to represent the influences of the large numbers of earthquake ground motions derived from the seismic ground motion prediction methods for the surrounding seismic sources. Viewing the relations between current design basis earthquake ground motion determination and modem earthquake ground motion estimation, a development of risk-informed design basis earthquake ground motion methodology is discussed for insight into the on going modernization of the Examination Guide for Seismic Design on NPP

  16. Design basis programs and improvements in plant operation

    International Nuclear Information System (INIS)

    Metcalf, M.F.

    1991-01-01

    Public Service Electric and Gas (PSE and G) Company operates three commercial nuclear power plants in southern New Jersey. The three plants are of different designs and vintages (two pressurized water reactors licensed in 1976 and 1980 and one boiling water reactor licensed in 1986). As the industry recognized the need to develop design basis programs, PSE and G also realized the need after a voluntary 52-day shutdown of one unit because of electrical design basis problems. In its drive to be a premier electric utility, PSE and G has been aggressively active in developing design basis documents (DBDs) with supporting projects and refined uses to obtain the expected value and see the return on investment. Progress on Salem is nearly 75% complete, while Hope Creek is 20% complete. To data, PSE and G has experienced success in the use of DBDs in areas such as development of plant modifications, development of the reliability-centered maintenance program, procedure upgrades, improved document retrieval, resolution of regulatory issues, and training. The paper examines the design basis development process, supporting projects, and expected improvements in plant operations as a result of these efforts

  17. Operation status and upgrading of HIRFL

    International Nuclear Information System (INIS)

    Tang, J.Y.; Wang, Y.F.; Wei, B.W.

    2001-01-01

    The operation status and the undergoing upgrading at HIRFL machine are presented. The accelerated ion species with the machine have been expanding, including metallic ions and higher energy with the new ECR ion source. The upgrading of HIRFL as the pre-accelerator of CSR storage ring has been processing steadily. The new 14.5 GHz ECR ion source has been put in operation in early 1999. A full-superconducting ECR ion source of 18 GHz is under design. The manufacture of the new vacuum chamber for SFC is just finished and the installation is to be started. The construction of the new B1 buncher is nearly to be finished, and the off-line test and the installation will be started soon. Another two identical bunchers will be ordered after the test. The beam distribution system is under upgrading to make all experiment stations separate from the others and the time-sharing mode possible, and a new cancer-therapy station is also under construction. The other upgrading items include the yoke enlarging of SFC, beam diagnostics, computer control and beam distribution system

  18. Nuclear safety improvement activities related to WWER-440 units in Bulgaria

    International Nuclear Information System (INIS)

    Gantchev, T.

    1998-01-01

    The systematic evaluation of the deficiencies of the original design of the WWER reactors brought to the development of a Short Term Programme for Safety Upgrading and Modernisation of Kozloduy WWER-440 units. The implementation of this Programme was completed in 1997. The strive for continuos improvement of Kozloduy Nuclear Power Plant (NPP) safety level, the new requirements of the Bulgarian Nuclear Safety Authority and the public concern initiated the development of new Complex Programme for Safety Improvement (PRG'97), now in a process of implementation. (author)

  19. Organic reactivity analysis in Hanford single-shell tanks: Experimental and modeling basis for an expanded safety criterion

    International Nuclear Information System (INIS)

    Fauske, H.; Grigsby, J.M.; Turner, D.A.; Babad, H.; Meacham, J.E.

    1996-01-01

    De-spite demonstrated safe storage in terms of chemical stability of the Hanford high level waste for many decades, including decreasing waste temperatures and continuing aging of chemicals to less energetic states, concerns continue relative to assurance of long-term safe storage. Review of potential chemical safety hazards has been of particular recent interest in response to serious incidents within the Nuclear Weapons Complexes in the former Soviet Union (the 1957 Kyshtym and the 1993 Tomsk-7 incidents). Based upon an evaluation of the extensive new information and understanding that have developed over the last few years, it is concluded that the Hanford waste is stored safely and that concerns related to potential chemical safety hazards are not warranted. Spontaneous bulk runaway reactions of the Kyshtym incident type and other potential condensed-phase propagating reactions can be ruled out by assuring appropriate tank operating controls are in place and by limiting tank intrusive activities. This paper summarizes the technical basis for this position

  20. Silicon Tracking Upgrade at CDF

    International Nuclear Information System (INIS)

    Kruse, M.C.

    1998-04-01

    The Collider Detector at Fermilab (CDF) is scheduled to begin recording data from Run II of the Fermilab Tevatron in early 2000. The silicon tracking upgrade constitutes both the upgrade to the CDF silicon vertex detector (SVX II) and the new Intermediate Silicon Layers (ISL) located at radii just beyond the SVX II. Here we review the design and prototyping of all aspects of these detectors including mechanical design, data acquisition, and a trigger based on silicon tracking

  1. Events as spaces for upgrading : Automotive events in Shanghai

    NARCIS (Netherlands)

    E. van Tuijl (Erwin); K. Dittrich (Koen)

    2015-01-01

    textabstractThis study contributes to the literature dealing with upgrading of the Chinese automotive industry by analysing the role of events in the upgrading process. By combining literature on temporary clusters with that of knowledge sourcing and upgrading, we investigate how firms use events

  2. Cost evaluation of I and C upgrade approach in nuclear power plants

    International Nuclear Information System (INIS)

    Kang, Hyun-Tae; Lee, Jae-Ki

    2013-01-01

    Highlights: • Cost evaluation process for I and C system upgrade is built. • 4 cost factors affecting I and C system upgrade are described. • Additional cost incurred by a phased upgrade is calculated. • Cost for system upgrade between upgrade implementations is compared. - Abstract: Utilities have recently been debating the respective pros and cons of implementation of a multi-phase upgrade during several normal outages versus a single major upgrade implementation during a prolonged outage. We have studied these approaches and have been developing the basic design of nuclear power plants (NPPs) instrumentation and control (I and C) upgrade since early 2008. As part of this study, analyses of the NPPs I and C systems were conducted and the need for upgrading the systems was raised. One of the primary concerns regarding the system upgrade is a cost-benefit implementation, which will influence the upgrade approach. From this viewpoint, the I and C upgrade must consider economic factors such as I and C vendor cost, architecture engineering cost, installation cost, utility cost, and other transition costs such as training and procedure development. This paper presents a comparison study of economical aspects including cost evaluation between the aforementioned upgrade implementations and suggests a solution for I and C upgrade approach

  3. Commissioning of the CMS Hadron Forward Calorimeters Phase I Upgrade

    CERN Document Server

    Bilki, Burak

    2017-01-01

    The final phase of the CMS Hadron Forward Calorimeters Phase I upgrade was performed during the Extended Year End Technical Stop of 2016 and 2017. In the framework of the upgrade, the PMT boxes were reworked to implement two channel readout in order to exploit the benefits of the multi-anode PMTs in background tagging and signal recovery. The front-end electronics were also upgraded to QIE10-based electronics which implement larger dynamic range and a 6-bit TDC. Following this major upgrade, the Hadron Forward Calorimeters were commissioned for operation readiness in 2017. Here we describe the details and the components of the upgrade, and discuss the operational experience and results obtained during the upgrade and commissioning.

  4. Commissioning of the CMS Hadron Forward Calorimeters Phase I Upgrade

    Science.gov (United States)

    Bilki, B.; Onel, Y.

    2018-03-01

    The final phase of the CMS Hadron Forward Calorimeters Phase I Upgrade was performed during the Extended Year End Technical Stop of 2016-2017. In the framework of the upgrade, the PMT boxes were reworked to implement two channel readout in order to exploit the benefits of the multi-anode PMTs in background tagging and signal recovery. The front-end electronics were also upgraded to QIE10-based electronics which implement larger dynamic range and a 6-bit TDC. Following this major upgrade, the Hadron Forward Calorimeters were commissioned for operation readiness in 2017. Here we describe the details and the components of the upgrade, and discuss the operational experience and results obtained during the upgrade and commissioning.

  5. LEP is upgraded

    CERN Multimedia

    1995-01-01

    A superconducting radio-frequency cavity is installed on the Large Electron-Positron (LEP) collider. This upgrade, known as LEP-2, allowed the accelerator to reach new, higher energies and so investigate new areas of physics.

  6. Habitat Demonstration Unit Medical Operations Workstation Upgrades

    Science.gov (United States)

    Trageser, Katherine H.

    2011-01-01

    This paper provides an overview of the design and fabrication associated with upgrades for the Medical Operations Workstation in the Habitat Demonstration Unit. The work spanned a ten week period. The upgrades will be used during the 2011 Desert Research and Technology Studies (Desert RATS) field campaign. Upgrades include a deployable privacy curtain system, a deployable tray table, an easily accessible biological waste container, reorganization and labeling of the medical supplies, and installation of a retractable camera. All of the items were completed within the ten week period.

  7. Get a winning Oracle upgrade session using the quarterback approach

    Science.gov (United States)

    Anderson, G.

    2002-01-01

    Upgrades, upgrades... too much customer down time. Find out how we shrunk our production upgrade schedule 40% from our estimate of 10 days 12 hours to 6 days 2 hours using the quarterback approach. So your upgrade is not that complex, come anyway. This approach is scalable to any size project and will be extremely valuable.

  8. PRELIMINARY SELECTION OF MGR DESIGN BASIS EVENTS

    International Nuclear Information System (INIS)

    Kappes, J.A.

    1999-01-01

    The purpose of this analysis is to identify the preliminary design basis events (DBEs) for consideration in the design of the Monitored Geologic Repository (MGR). For external events and natural phenomena (e.g., earthquake), the objective is to identify those initiating events that the MGR will be designed to withstand. Design criteria will ensure that radiological release scenarios resulting from these initiating events are beyond design basis (i.e., have a scenario frequency less than once per million years). For internal (i.e., human-induced and random equipment failures) events, the objective is to identify credible event sequences that result in bounding radiological releases. These sequences will be used to establish the design basis criteria for MGR structures, systems, and components (SSCs) design basis criteria in order to prevent or mitigate radiological releases. The safety strategy presented in this analysis for preventing or mitigating DBEs is based on the preclosure safety strategy outlined in ''Strategy to Mitigate Preclosure Offsite Exposure'' (CRWMS M andO 1998f). DBE analysis is necessary to provide feedback and requirements to the design process, and also to demonstrate compliance with proposed 10 CFR 63 (Dyer 1999b) requirements. DBE analysis is also required to identify and classify the SSCs that are important to safety (ITS)

  9. Slum upgrading in developing countries: lessons from Ghana and ...

    African Journals Online (AJOL)

    The paper identifies common elements in the upgrading programmes in the literature. It assesses two slum upgrading projects from Ghana and Kenya to determine how the elements were factored into the projects' implementation. The article concludes that stakeholders involved in slum upgrading in Africa should consider ...

  10. 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.

  11. Promoting safety in nuclear installations. The IAEA has established safety standards for nuclear reactors and provides expert review and safety services to assist Member States in their application

    International Nuclear Information System (INIS)

    2002-01-01

    More than 430 nuclear power plants (NPPs) are currently operating in 30 countries around the world. The nuclear share of total electricity production ranges from about 20 percent in the Czech Republic and United States to nearly 78 percent in France and Lithuania. Worldwide, nuclear power generates about 16% of the total electricity. The safety of such nuclear installations is fundamental. Every aspect of a power plant must be closely supervised and scrutinized by national regulatory bodies to ensure safety at every phase. These aspects include design, construction, commissioning, trial operation, commercial operation, repair and maintenance, plant upgrades, radiation doses to workers, radioactive waste management and, ultimately, plant decommissioning. Safety fundamentals comprise defence-in-depth, which means having in place multiple levels of protection. nuclear facilities; regulatory responsibility; communicating with the public; adoption of the international convention on nuclear safety including implementation of IAEA nuclear safety standards. This publication covers topics of designing for safety (including safety concepts, design principles, and human factors); operating safety (including safety culture and advance in operational safety); risk assessment and management

  12. The cohort of the atomic bomb survivors major basis of radiation safety regulations

    CERN Document Server

    Rühm, W; Nekolla, E A

    2006-01-01

    Since 1950 about 87 000 A-bomb survivors from Hiroshima and Nagasaki have been monitored within the framework of the Life Span Study, to quantify radiation-induced late effects. In terms of incidence and mortality, a statistically significant excess was found for leukemia and solid tumors. In another major international effort, neutron and gamma radiation doses were estimated, for those survivors (Dosimetry System DS02). Both studies combined allow the deduction of risk coefficients that serve as a basis for international safety regulations. As an example, current results on all solid tumors combined suggest an excess relative risk of 0.47 per Sievert for an attained age of 70 years, for those who were exposed at an age of 30 years. After exposure to an effective dose of one Sievert the solid tumor mortality would thus be about 50% larger than that expected for a similar cohort not exposed to any ionizing radiation from the bombs.

  13. The LHCb VELO upgrade

    International Nuclear Information System (INIS)

    Rodríguez Pérez, Pablo

    2013-01-01

    LHCb is a forward spectrometer experiment dedicated to the study of new physics in the decays of beauty and charm hadrons produced in proton collisions at the Large Hadron Collider (LHC) at CERN. The VErtex LOcator (VELO) is the microstrip silicon detector surrounding the interaction point, providing tracking and vertexing measurements. The upgrade of the LHCb experiment, planned for 2018, will increase the luminosity up to 2×10 33 cm −2 s −1 and will perform the readout as a trigger-less system with an event rate of 40 MHz. Extremely non-uniform radiation doses will reach up to 5×10 15 1 MeV n eq /cm 2 in the innermost regions of the VELO sensors, and the output data bandwidth will be increased by a factor of 40. An upgraded detector is under development based in a pixel sensor of the Timepix/Medipix family, with 55×55μm 2 pixels. In addition a microstrip solution with finer pitch, higher granularity and thinner than the current detector is being developed in parallel. The current status of the VELO upgrade program will be described together with recent testbeam results

  14. Qinshan 300Mwe NPP full scope simulator upgrade

    International Nuclear Information System (INIS)

    Qi Kelin; Li Qing; Liu Wei, Lai Shengyuan

    2006-01-01

    On April 28,2004, RINPO was awarded the project for Qinshan 300Mwe NPP full scope simulator upgrade, the SAT (site acceptance test) was completed on June 30 2005 and the simulator put into operator training again. Scope of upgrade includes: computer system (DGI server and workstations) all replaced by microcomputers; G2 I/O controllers all replaced by RTP EIOBC; Unix-based simulation support environment replaced by RINPO's PC-based simulation environment RINSIMTM, Instructor software replaced by RINPO's PC-based instructor software with function and diagram redesigned; DEH, Feed-water control and some other digital control systems redeveloped to follow NPP modifications; desk-top simulator with soft panel control room developed as byproduct; most of the models not changed but it is planned the reactor core and PPC model will be upgraded in near future. SAT of upgrade demonstrates that the performance of the simulator much improved after the upgrade. (author)

  15. Coal upgrading

    Energy Technology Data Exchange (ETDEWEB)

    Nunes, S. [IEA Clean Coal Centre, London (United Kingdom)

    2009-10-15

    This report examines current technologies and those likely to be used to produce cleaner coal and coal products, principally for use in power generation and metallurgical applications. Consideration is also given to coal production in the leading coal producing countries, both with developed and developing industries. A range of technologies are considered. These include the coal-based liquid fuel called coal water mixture (CWM) that may compete with diesel, the production of ultra-clean coal (UCC) and coal liquefaction which competes with oil and its products. Technologies for upgrading coal are considered, especially for low rank coals (LRC), since these have the potential to fill the gap generated by the increasing demand for coal that cannot be met by higher quality coals. Potential advantages and downsides of coal upgrading are outlined. Taking into account the environmental benefits of reduced pollution achieved through cleaner coal and reduced transport costs, as well as other positive aspects such as a predictable product leading to better boiler design, the advantages appear to be significant. The drying of low rank coals improves the energy productively released during combustion and may also be used as an adjunct or as part of other coal processing procedures. Coal washing technologies vary in different countries and the implications of this are outlined. Dry separation technologies, such as dry jigging and electrostatic separation, are also described. The demonstration of new technologies is key to their further development and demonstrations of various clean coal technologies are considered. A number of approaches to briquetting and pelletising are available and their use varies from country to country. Finally, developments in upgrading low rank coals are described in the leading coal producing countries. This is an area that is developing rapidly and in which there are significant corporate and state players. 81 refs., 32 figs., 3 tabs.

  16. Upgrade of of monitoring devices for radiation information

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Taek; Park, Won Man; Kim, Jung Soo; Lee, Bong Jae; Jae Yoo Kyung

    1999-01-01

    The Radiation Monitoring System (RMS) in Hanaro of KAERI and NPPs of Korea, supplied by victoreen, sometimes has been stopped to use slow 486 PC (Personal computer) and not to be enough memory in main processing computer, IOCA and IOCB, for signal processing and storing. It is very difficult for operator to operate and maintain RMS, because of using an unfamiliar operating system, SCO Unix, of main computer. And also, ScanRad (TM) program for processing and storing radiation signal has Y2K problems and is able to lose and not to display measuring signals. Therefore it needs to upgrade the computer system in RMS. This study is upgrading the main computer, IOCA, in RMS of Hanaro to Pentium PC, and changing operating System to Window NT-based system. Therefore it needs to upgrade the computer system in RMS. This study is upgrading the main computer, IOCA, in RMS of Hanaro to Pentium PC, and changing Operating System to Window NT-based system. Therefore a performance of the computer system in RMS has been upgraded for operator to be useful. This study is going two steps. First, the main computer, IOCA a part of the whole computer system has been upgraded to Pentium PC, and changed to Window NT-based system. Second, all of the computer system in Hanaro RMS is going to be upgraded. This study has got the following results: a RS-232C serial communication program: between the upgraded IOCA and LCU (Local Control Unit) -- a serial communication test configurating two LCU serial --a parallel communication test configurating two LCU parallel: GUI program to present a radioactive information -- overview schematic display page -- detail display pages -- alarm and event pages -- trend pages and group trend pages.

  17. Upgrade of of monitoring devices for radiation information

    International Nuclear Information System (INIS)

    Kim, Jung Taek; Park, Won Man; Kim, Jung Soo; Lee, Bong Jae; Jae Yoo Kyung

    1999-01-01

    The Radiation Monitoring System (RMS) in Hanaro of KAERI and NPPs of Korea, supplied by victoreen, sometimes has been stopped to use slow 486 PC (Personal computer) and not to be enough memory in main processing computer, IOCA and IOCB, for signal processing and storing. It is very difficult for operator to operate and maintain RMS, because of using an unfamiliar operating system, SCO Unix, of main computer. And also, ScanRad (TM) program for processing and storing radiation signal has Y2K problems and is able to lose and not to display measuring signals. Therefore it needs to upgrade the computer system in RMS. This study is upgrading the main computer, IOCA, in RMS of Hanaro to Pentium PC, and changing operating System to Window NT-based system. Therefore it needs to upgrade the computer system in RMS. This study is upgrading the main computer, IOCA, in RMS of Hanaro to Pentium PC, and changing Operating System to Window NT-based system. Therefore a performance of the computer system in RMS has been upgraded for operator to be useful. This study is going two steps. First, the main computer, IOCA a part of the whole computer system has been upgraded to Pentium PC, and changed to Window NT-based system. Second, all of the computer system in Hanaro RMS is going to be upgraded. This study has got the following results: a RS-232C serial communication program: between the upgraded IOCA and LCU (Local Control Unit) -- a serial communication test configurating two LCU serial --a parallel communication test configurating two LCU parallel: GUI program to present a radioactive information -- overview schematic display page -- detail display pages -- alarm and event pages -- trend pages and group trend pages

  18. The UKIRT Upgrades Programme

    Science.gov (United States)

    Adamson, Andy; Davies, John; Robson, Ian

    Tim Hawarden presented this paper to the 30th anniversary workshop, just a month before his untimely death. The editors have done their best to convert his talk into this paper, and gratefully acknowledge the assistance of Nick Rees (a member of the Upgrades team, now at Diamond Light Source). Tim's discussion concerned the UKIRT Upgrades Project, which ran through the 1990s and transformed the telescope and made it truly competitive on the world stage for operation into the twenty-first century. The reference list at the end of the paper is comprehensive; some of these are referred to in the paper itself and some are included for completeness only.

  19. Reflections on a digital upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Tadjalli, M.

    2013-07-01

    Upon receiving US NRC's approval in 2010, the first RPS/ESPS digital upgrade using TELEPERM® XS technology was successfully installed in Spring of 2011 at the first Unit of a three-unit station, followed by the 2nd Unit installation in spring of 2012. Both Units' systems have been operating flawlessly since installation. After about two years of operation, a reflection on digital upgrades and lessons learned, from a vendor perspective, provides valuable insight for the commercial nuclear power industry.

  20. Argonne's atlas control system upgrade

    International Nuclear Information System (INIS)

    Munson, F.; Quock, D.; Chapin, B.; Figueroa, J.

    1999-01-01

    The ATLAS facility (Argonne Tandem-Linac Accelerator System) is located at the Argonne National Laboratory. The facility is a tool used in nuclear and atomic physics research, which focuses primarily on heavy-ion physics. The accelerator as well as its control system are evolutionary in nature, and consequently, continue to advance. In 1998 the most recent project to upgrade the ATLAS control system was completed. This paper briefly reviews the upgrade, and summarizes the configuration and features of the resulting control system

  1. US utility I ampersand C upgrade initiative

    International Nuclear Information System (INIS)

    Wilkinson, D.

    1993-01-01

    This paper describes an initiative which is being pursued to address the problems associated with the increasing obsolescence of instrumentation and control (I ampersand C) systems in nuclear power plants. Such efforts will improve safety, availability, and reliability of plant operation. The program presented is a joint plan between EPRI, utilities, and vendors, to bring about the development of technology to meet these needs, to encourage the private sector to develop such technology, and to prove to utilities the value and utility in applying such technology to upgrade their facilities. Part of this plan is demonstration programs where plant-wide sensor and I ampersand C improvements can be implemented, in a documented manner, to provide R ampersand D information on their performance, functionality, and reliability. The intent is to reduce the costs and shorten the learning curve associated with the utilization of modern I ampersand C equipment

  2. Completion of the ATLAS control system upgrade

    International Nuclear Information System (INIS)

    Munson, F. H.

    1998-01-01

    In the fall of 1992 at the SNEAP(Symposium of North Eastern Accelerator Personnel) a project to up grade the ATLAS (Argonne Tandem Linear Accelerator System) control system was first reported. Not unlike the accelerator it services the control system will continue to evolve. However, the first of this year has marked the completion of this most recent upgrade project. Since the control system upgrade took place during a period when ATLAS was operating at a record number of hours, special techniques were necessary to enable the development of the new control system ''on line'' while still saving the needs of normal operations. This paper reviews the techniques used for upgrading the ATLAS control system while the system was in use. In addition a summary of the upgrade project and final configuration, as well as some of the features of the new control system is provided

  3. A Collaborative Webbased Framework with Optimized Mobile Synchronisation: Upgrading to Medicine 2.0

    Directory of Open Access Journals (Sweden)

    P.L. Kubben

    2008-07-01

    Full Text Available Magazines like BusinessWeek and Time haveconfirmed “The Power of Us” as we collaborate in the Web2.0. However, current medical practice still relies onMedicine 1.0. The use of webbased collaborativeframeworks (“wiki’s” combined with optimized output andsynchronisation for mobile devices can help us upgrading toMedicine 2.0. This serves as an aid for community-drivenclinical decision support systems increasing medical safety.NeuroWiki.com will be demonstrated as a useful concept inthis philosophy, and its technical background will bedescribed.

  4. Is your upgrade worth it? process mining can tell

    NARCIS (Netherlands)

    Genuchten, van M.J.I.M.; Mans, R.S.; Reijers, H.A.; Wismeijer, D.

    2014-01-01

    Software vendors typically release updates and upgrades of their software once or twice a year. Users are then faced with the question of whether the upgrade is worth the price and the trouble. The software industry doesn't provide much evidence that it's worthwhile to upgrade to new releases. The

  5. Is your upgrade worth it? Process mining can tell

    NARCIS (Netherlands)

    Genuchten, M.J.I.M.; Mans, R.S.; Reijers, H.A.; Wismeijer, D.

    2014-01-01

    Software vendors typically release updates and upgrades of their software once or twice a year. Users are then faced with the question of whether the upgrade is worth the price and the trouble. The software industry doesn't provide much evidence that it's worthwhile to upgrade to new releases. The

  6. Scenarios for the LHC Upgrade

    CERN Document Server

    Scandale, Walter

    2008-01-01

    The projected lifetime of the LHC low-beta quadrupoles, the evolution of the statistical error halving time, and the physics potential all call for an LHC luminosity upgrade by the middle of the coming decade. In the framework of the CARE-HHH network three principal scenarios have been developed for increasing the LHC peak luminosity by more than a factor of 10, to values above 1035 cm−2s−1. All scenarios imply a rebuilding of the high-luminosity interaction regions (IRs) in combination with a consistent change of beam parameters. However, their respective features, bunch structures, IR layouts, merits and challenges, and luminosity variation with β∗ differ substantially. In all scenarios luminosity leveling during a store would be advantageous for the physics experiments. An injector upgrade must complement the upgrade measures in the LHC proper in order to provide the beam intensity and brightness needed as well as to reduce the LHC turnaround time for higher integrated luminosity.

  7. Hydrogen assisted biological biogas upgrading

    DEFF Research Database (Denmark)

    Bassani, Ilaria

    Wind and biomass are promoted worldwide as sustainable forms of energy. Anaerobic digestion of biomass produces biogas with ∼50−70% CH4 and 30−50% CO2. However, biogas with >90% CH4 content has higher heating value, can be injected into the natural gas grid or used as alternative to natural gas...... as vehicle fuel. Methods currently available for biogas upgrading mainly consists of physicochemical CO2 removal, requiring the use of chemical substances and energy input and, thus, increasing process costs. This PhD project proposes an alternative to existing biogas upgrading technologies, where H2......, produced by water electrolysis, using excess of electricity from wind mills, is coupled with the CO2 contained in the biogas to convert them to CH4. This process is defined as biological biogas upgrading and is carried out by hydrogenotrophic methanogenic archaea that couples CO2 with H2 to produce...

  8. Conception of transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel of power reactors, which meets the requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism

    International Nuclear Information System (INIS)

    Il'kaev, R.I.; Matveev, V.Z.; Morenko, A.I.; Shapovalov, V.I.; Semenov, A.G.; Sergeyev, V.M.; Orlov, V.K.; Shatalov, V.V.; Gotovchikov, V.T.; Seredenko, V.A.; Haire, Jonathan M.; Forsberg, C.W.

    2004-01-01

    The report is devoted to the problem of creation of a new generation of multi-purpose universal transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel (SNF) of power reactors, which meets all requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism. Meeting all IAEA requirements in terms of safety both in normal operation conditions and accidents, as well as increased stability of transport cask (TC) with SNF under the conditions of beyond-design-basis accidents and acts of terrorism has been achieved in the design of multi-purpose universal TC due to the use of DU (depleted uranium) in it. At that, it is suggested to use DU in TC, which acts as effective gamma shield and constructional material in the form of both metallic depleted uranium and metal-ceramic mixture (cermet), based on stainless or carbon steel and DU dioxide. The metal in the cermet is chosen to optimize cask performance. The use of DU in the design of multi-purpose universal TC enables getting maximum load of the container for spent nuclear fuel when meeting IAEA requirements in terms of safety and providing increased stability of the container with SNF under conditions of beyond-design-basis accident and acts of terrorism

  9. Conception of transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel of power reactors, which meets the requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism

    Energy Technology Data Exchange (ETDEWEB)

    Il' kaev, R.I.; Matveev, V.Z.; Morenko, A.I.; Shapovalov, V.I. [Russian Federal Nuclear Center - All-Russian Research Inst. of Experimental Physics, Sarov (Russian Federation); Semenov, A.G.; Sergeyev, V.M.; Orlov, V.K. [All-Russian Research Inst. of Inorganic Materials, Moscow (Russian Federation); Shatalov, V.V.; Gotovchikov, V.T.; Seredenko, V.A. [All-Russian Research Inst. of Applied Chemistry, Moscow (Russian Federation); Haire, Jonathan M.; Forsberg, C.W. [Oak Ridge National Lab., Oak Ridge (United States)

    2004-07-01

    The report is devoted to the problem of creation of a new generation of multi-purpose universal transport cask with advanced safety, aimed at transportation and storage of spent nuclear fuel (SNF) of power reactors, which meets all requirements of IAEA in terms of safety and increased stability during beyond-design-basis accidents and acts of terrorism. Meeting all IAEA requirements in terms of safety both in normal operation conditions and accidents, as well as increased stability of transport cask (TC) with SNF under the conditions of beyond-design-basis accidents and acts of terrorism has been achieved in the design of multi-purpose universal TC due to the use of DU (depleted uranium) in it. At that, it is suggested to use DU in TC, which acts as effective gamma shield and constructional material in the form of both metallic depleted uranium and metal-ceramic mixture (cermet), based on stainless or carbon steel and DU dioxide. The metal in the cermet is chosen to optimize cask performance. The use of DU in the design of multi-purpose universal TC enables getting maximum load of the container for spent nuclear fuel when meeting IAEA requirements in terms of safety and providing increased stability of the container with SNF under conditions of beyond-design-basis accident and acts of terrorism.

  10. Status of the Upgrade of the CERN PS Booster

    CERN Document Server

    Hanke, K; Angoletta, M; Bartmann, W; Bartolome, S; Benedetto, E; Bertone, C; Blas, A; Bonnal, P; Borburgh, J; Bozzini, D; Butterworth, A; Carli, C; Carlier, E; Cole, J; Dahlen, P; Delonca, M; Dobers, T; Findlay, A; Froeschl, R; Hansen, J; Hay, D; Jensen, S; Lacroix, J; Le Roux, P; Lopez Hernandez, L; Maglioni, C; Masi, A; Mason, G; Mathot, S; Mikulec, B; Muttoni, Y; Newborough, A; Nisbet, D; Olek, S; Paoluzzi, M; Perillo-Marcone, A; Pittet, S; Puccio, B; Raginel, V; Riffaud, B; Ruehl, I; Sarrió Martínez, A; Tan, J; Todd, B; Venturi, V; Weterings, W

    2013-01-01

    The CERN PS Booster (PSB) is presently undergoing an ambitious consolidation and upgrade program within the frame of the LHC Injectors Upgrade (LIU) project. This program comprises a new injection scheme for H- ions from CERN’s new Linac4, the replacement of the main RF systems and an energy upgrade of the PSB rings from 1.4 to 2 GeV which includes the replacement of the main magnet power supply as well as the upgrade of the extraction equipment. This paper describes the status and plans of this work program.

  11. The Upgrade to Hybrid Incubators in China

    DEFF Research Database (Denmark)

    Hu, Yimei; Gao, Yuchen

    countries and emerging economics. Based on a single in-depth case-study of Tuspark Incubator, this study explores key factors and specific ways for the upgrade to hybrid incubators in the context of China. By using categorical analysis, three factors, i.e. incubation subdivision, intermediary platform......, and proactive approach, are found to be essential for a formerly government-sponsored TBI’s upgrading. The result of this study also provides new insights and several implications for incubator managers and policy makers in emerging economies. In addition, whether the key factors can be used in upgrade of other...... TBIs in China requires further study....

  12. Nuclear safety activities in the SR of Slovenia in 1986

    Energy Technology Data Exchange (ETDEWEB)

    Susnik, J [Inst. Jozef Stefan, Ljubljana (Slovenia)

    1987-06-15

    Currently Yugoslavia has one 632 MWe nuclear power plant (NPP) of PWR design, located at Krsko in the Socialist Republic (SR) of Slovenia. Krsko NPP, which is a two-loop plant, started power operation in 1981. In general, reactor safety activities in the SR of Slovenia are mostly related to upgrading the safety of our Krsko NPP and to developing capabilities for use in future units. This report presents the nuclear safety related legislation and organization of the corresponding regulatory body, and the activities related to nuclear safety of the participating organizations in the SR of Slovenia in 1986. (author)

  13. Nuclear safety activities in the SR of Slovenia in 1986

    International Nuclear Information System (INIS)

    Susnik, J.

    1987-06-01

    Currently Yugoslavia has one 632 MWe nuclear power plant (NPP) of PWR design, located at Krsko in the Socialist Republic (SR) of Slovenia. Krsko NPP, which is a two-loop plant, started power operation in 1981. In general, reactor safety activities in the SR of Slovenia are mostly related to upgrading the safety of our Krsko NPP and to developing capabilities for use in future units. This report presents the nuclear safety related legislation and organization of the corresponding regulatory body, and the activities related to nuclear safety of the participating organizations in the SR of Slovenia in 1986. (author)

  14. Commissioning of the CMS Hadron Forward Calorimeters Phase I Upgrade

    CERN Document Server

    Bilki, Burak

    2017-01-01

    The final phase of the CMS Hadron Forward Calorimeters Phase I upgrade is being performed during the Extended Year End Technical Stop of 2016 â?? 2017. In the framework of the upgrade, the PMT boxes are being reworked to implement two channel readout in order to exploit the benefits of the multi-anode PMTs in background tagging and signal recovery. The front-end electronics is also being upgraded to QIE10-based electronics which will implement larger dynamic range and a 6-bit TDC to eliminate the background to have an effect on the trigger. Following this major upgrade, the Hadron Forward Calorimeters will be commissioned for operation readiness in 2017. Here we describe the details and the components of the upgrade, and discuss the operational experience and results obtained during the upgrade and commissioning.

  15. Canister storage building design basis accident analysis documentation

    International Nuclear Information System (INIS)

    KOPELIC, S.D.

    1999-01-01

    This document provides the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report

  16. Quality assurance requirements for the computer software and safety analyses

    International Nuclear Information System (INIS)

    Husarecek, J.

    1992-01-01

    The requirements are given as placed on the development, procurement, maintenance, and application of software for the creation or processing of data during the design, construction, operation, repair, maintenance and safety-related upgrading of nuclear power plants. The verification and validation processes are highlighted, and the requirements put on the software documentation are outlined. The general quality assurance principles applied to safety analyses are characterized. (J.B.). 1 ref

  17. A cross-tokamak neural network disruption predictor for the JET and ASDEX Upgrade tokamaks

    Science.gov (United States)

    Windsor, C. G.; Pautasso, G.; Tichmann, C.; Buttery, R. J.; Hender, T. C.; EFDA Contributors, JET; ASDEX Upgrade Team

    2005-05-01

    First results are reported on the prediction of disruptions in one tokamak, based on neural networks trained on another tokamak. The studies use data from the JET and ASDEX Upgrade devices, with a neural network trained on just seven normalized plasma parameters. In this way, a simple single layer perceptron network trained solely on JET correctly anticipated 67% of disruptions on ASDEX Upgrade in advance of 0.01 s before the disruption. The converse test led to a 69% success rate in advance of 0.04 s before the disruption in JET. Only one overall time scaling parameter is allowed between the devices, which can be introduced from theoretical arguments. Disruption prediction performance based on such networks trained and tested on the same device shows even higher success rates (JET, 86%; ASDEX Upgrade, 90%), despite the small number of inputs used and simplicity of the network. It is found that while performance for networks trained and tested on the same device can be improved with more complex networks and many adjustable weights, for cross-machine testing the best approach is a simple single layer perceptron. This offers the basis of a potentially useful technique for large future devices such as ITER, which with further development might help to reduce disruption frequency and minimize the need for a large disruption campaign to train disruption avoidance systems.

  18. A cross-tokamak neural network disruption predictor for the JET and ASDEX Upgrade tokamaks

    International Nuclear Information System (INIS)

    Windsor, C.G.; Buttery, R.J.; Hender, T.C.; Pautasso, G.; Tichmann, C.

    2005-01-01

    First results are reported on the prediction of disruptions in one tokamak, based on neural networks trained on another tokamak. The studies use data from the JET and ASDEX Upgrade devices, with a neural network trained on just seven normalized plasma parameters. In this way, a simple single layer perceptron network trained solely on JET correctly anticipated 67% of disruptions on ASDEX Upgrade in advance of 0.01 s before the disruption. The converse test led to a 69% success rate in advance of 0.04 s before the disruption in JET. Only one overall time scaling parameter is allowed between the devices, which can be introduced from theoretical arguments. Disruption prediction performance based on such networks trained and tested on the same device shows even higher success rates (JET, 86%; ASDEX Upgrade, 90%), despite the small number of inputs used and simplicity of the network. It is found that while performance for networks trained and tested on the same device can be improved with more complex networks and many adjustable weights, for cross-machine testing the best approach is a simple single layer perceptron. This offers the basis of a potentially useful technique for large future devices such as ITER, which with further development might help to reduce disruption frequency and minimize the need for a large disruption campaign to train disruption avoidance systems

  19. Model review and evaluation for application in DOE safety basis documentation of chemical accidents - modeling guidance for atmospheric dispersion and consequence assessment

    Energy Technology Data Exchange (ETDEWEB)

    Lazaro, M. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Woodarad, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Hanna, S. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hesse, D. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Huang, J. -C. [Argonne National Lab. (ANL), Argonne, IL (United States); Lewis, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Mazzola, C. A. [Argonne National Lab. (ANL), Argonne, IL (United States)

    1997-09-01

    The U.S. Department of Energy (DOE), through its Defense Programs (DP), Office of Engineering and Operations Suppon, established the Accident Phenomenology and Consequence (AP AC) Methodology Evaluation Program to identify and evaluate methodologies and computer codes to support accident phenomenological and consequence calculations for both radiological and nonradiological materials at DOE facilities and to identify development needs. The program is also intended to define and recommend "best or good engineering/safety analysis practices" to be followed in preparing ''design or beyond design basis" assessments to be included in DOE nuclear and nonnuclear facility safety documents. The AP AC effort is intended to provide scientifically sound and more consistent analytical approaches, by identifying model selection procedures and application methodologies, in order to enhance safety analysis activities throughout the DOE complex.

  20. Optimization bitumen-based upgrading and refining schemes

    Energy Technology Data Exchange (ETDEWEB)

    Munteanu, M.; Chen, J. [National Centre for Upgrading Technology, Devon, AB (Canada); Natural Resources Canada, Devon, AB (Canada). CanmetENERGY

    2009-07-01

    This poster highlighted the results of a study in which the entire refining scheme for Canadian bitumen as feedstocks was modelled and simulated under different process configurations, operating conditions and product structures. The aim of the study was to optimize the economic benefits, product quality and energy use under a range of operational scenarios. Optimal refining schemes were proposed along with process conditions for existing refinery configurations and objectives. The goal was to provide guidelines and information for upgrading and refining process design and retrofitting. Critical steps were identified with regards to the upgrading process. It was concluded that the information obtained from this study would lead to significant improvement in process performance and operations, and in reducing the capital cost for building new upgraders and refineries. The simulation results provided valuable information for increasing the marketability of bitumen, reducing greenhouse gas emissions and other environmental impacts associated with bitumen upgrading and refining. tabs., figs.

  1. Instrument hardware and software upgrades at IPNS

    International Nuclear Information System (INIS)

    Worlton, Thomas; Hammonds, John; Mikkelson, D.; Mikkelson, Ruth; Porter, Rodney; Tao, Julian; Chatterjee, Alok

    2006-01-01

    IPNS is in the process of upgrading their time-of-flight neutron scattering instruments with improved hardware and software. The hardware upgrades include replacing old VAX Qbus and Multibus-based data acquisition systems with new systems based on VXI and VME. Hardware upgrades also include expanded detector banks and new detector electronics. Old VAX Fortran-based data acquisition and analysis software is being replaced with new software as part of the ISAW project. ISAW is written in Java for ease of development and portability, and is now used routinely for data visualization, reduction, and analysis on all upgraded instruments. ISAW provides the ability to process and visualize the data from thousands of detector pixels, each having thousands of time channels. These operations can be done interactively through a familiar graphical user interface or automatically through simple scripts. Scripts and operators provided by end users are automatically included in the ISAW menu structure, along with those distributed with ISAW, when the application is started

  2. Safety evaluation report of the Waste Isolation Pilot Plant safety analysis report: Contact-handled transuranic waste disposal operations

    International Nuclear Information System (INIS)

    1997-02-01

    DOE 5480.23, Nuclear Safety Analysis Reports, requires that the US Department of Energy conduct an independent, defensible, review in order to approve a Safety Analysis Report (SAR). That review and the SAR approval basis is documented in this formal Safety Evaluation Report (SER). This SER documents the DOE's review of the Waste Isolation Pilot Plant SAR and provides the Carlsbad Area Office Manager, the WIPP SAR approval authority, with the basis for approving the safety document. It concludes that the safety basis documented in the WIPP SAR is comprehensive, correct, and commensurate with hazards associated with planned waste disposal operations

  3. Methods used to seismically upgrade. The safety related components of Belgian plants

    International Nuclear Information System (INIS)

    Lafaille, J.P.

    1993-01-01

    Belgian nuclear power amounts to about 6,000 MW, generated by seven plants that started operation as early as 1967. The latest plant started in 1985. Some of these plants were designed with no seismic requirements whatsoever. Even for those that had seismic requirements at the design stage, seismic demand was raised after design had been frozen (late during construction or at the 10 years revision). As a consequence all the plants had to undergo, to a variable extent, a seismic reevaluation and/or backfitting. Civil structures were concerned as well as electro-mechanical equipment and piping systems. The present paper deals with the mechanical aspect of the problem (equipment and piping). In order to minimize hardware modifications, advanced analytical techniques were used throughout the process, starting with the elaboration of a site specific spectrum, and using a full soil-structure interaction in order to get as 'realistic' as possible floor response spectra. In some instances, non linear elasto-plastic time history analysis was performed on piping-systems in order to qualify them without hardware modifications. In other cases a 'Load Coefficient Method' was used. Sometimes stresses or displacements taken from the original stress reports and scaled by comparison of applicable spectra, allowed to assess the seismic validity of the system under investigation. Seismic acceptability of installed active equipment is more difficult to demonstrate, as this is usually done by testing. This problem is a generic issue in the US, identified under the label USI-A-46 (Unresolved Safety Issue). It is treated by. a group of Utilities (SQUG = Seismic Qualification Utilities Group). The Belgian Utility is member of that group since 1985. The application of this program is starting in the US. SQUG methodology has been applied to three Belgian plants starting in 1988 and is now completed. The required fixes are being implemented. Experience gained in the process has been applied

  4. Human-machine interface upgrade

    International Nuclear Information System (INIS)

    Kropik, M.; Matejka, K.; Sklenka, L.; Chab, V.

    2002-01-01

    The article describes a new human-machine interface that was installed at the VR-1 training reactor. The human-machine interface upgrade was completed in the summer 2001. The interface was designed with respect to functional, ergonomic and aesthetic requirements. The interface is based on a personal computer equipped with two displays. One display enables alphanumeric communication between the reactor operator and the nuclear reactor I and C. The second display is a graphical one. It presents the status of the reactor, principal parameters (as power, period), control rods positions, course of the reactor power. Furthermore, it is possible to set parameters, to show the active core configuration, to perform reactivity calculations, etc. The software for the new human-machine interface was produced with the InTouch developing tool of the Wonder-Ware Company. It is possible to switch the language of the interface between Czech and English because of many foreign students and visitors to the reactor. Microcomputer based communication units with proper software were developed to connect the new human-machine interface with the present reactor I and C. The new human-machine interface at the VR-1 training reactor improves the comfort and safety of the reactor utilisation, facilitates experiments and training, and provides better support for foreign visitors. (orig.)

  5. Development of requirements for seismic upgrading of equipment of existing WWER-440 and WWER-1000 type NPPs

    International Nuclear Information System (INIS)

    Kaznovsky, S.; Ostretsov, I.

    1993-01-01

    The change in seismology data and safety demands a necessity arose for seismic upgrading of the existing operating NPPs of WWER type which have been originally designed and built without or with simplifies calculations of seismic influences. The paper describes the traditional methods and approaches and calculation-experimental method for examining and ensuring of equipment seismic resistance at the NPPs directly. Method of ground explosions is included as well

  6. Upgrades of DARWIN, a dose and spectrum monitoring system applicable to various types of radiation over wide energy ranges

    Science.gov (United States)

    Sato, Tatsuhiko; Satoh, Daiki; Endo, Akira; Shigyo, Nobuhiro; Watanabe, Fusao; Sakurai, Hiroki; Arai, Yoichi

    2011-05-01

    A dose and spectrum monitoring system applicable to neutrons, photons and muons over wide ranges of energy, designated as DARWIN, has been developed for radiological protection in high-energy accelerator facilities. DARWIN consists of a phoswitch-type scintillation detector, a data-acquisition (DAQ) module for digital waveform analysis, and a personal computer equipped with a graphical-user-interface (GUI) program for controlling the system. The system was recently upgraded by introducing an original DAQ module based on a field programmable gate array, FPGA, and also by adding a function for estimating neutron and photon spectra based on an unfolding technique without requiring any specific scientific background of the user. The performance of the upgraded DARWIN was examined in various radiation fields, including an operational field in J-PARC. The experiments revealed that the dose rates and spectra measured by the upgraded DARWIN are quite reasonable, even in radiation fields with peak structures in terms of both spectrum and time variation. These results clearly demonstrate the usefulness of DARWIN for improving radiation safety in high-energy accelerator facilities.

  7. Upgrades of DARWIN, a dose and spectrum monitoring system applicable to various types of radiation over wide energy ranges

    International Nuclear Information System (INIS)

    Sato, Tatsuhiko; Satoh, Daiki; Endo, Akira; Shigyo, Nobuhiro; Watanabe, Fusao; Sakurai, Hiroki; Arai, Yoichi

    2011-01-01

    A dose and spectrum monitoring system applicable to neutrons, photons and muons over wide ranges of energy, designated as DARWIN, has been developed for radiological protection in high-energy accelerator facilities. DARWIN consists of a phoswitch-type scintillation detector, a data-acquisition (DAQ) module for digital waveform analysis, and a personal computer equipped with a graphical-user-interface (GUI) program for controlling the system. The system was recently upgraded by introducing an original DAQ module based on a field programmable gate array, FPGA, and also by adding a function for estimating neutron and photon spectra based on an unfolding technique without requiring any specific scientific background of the user. The performance of the upgraded DARWIN was examined in various radiation fields, including an operational field in J-PARC. The experiments revealed that the dose rates and spectra measured by the upgraded DARWIN are quite reasonable, even in radiation fields with peak structures in terms of both spectrum and time variation. These results clearly demonstrate the usefulness of DARWIN for improving radiation safety in high-energy accelerator facilities.

  8. Electrical and mechanical adaptation of commercially available power inverter modules for BUSSARD – The power supply of ASDEX Upgrade in vessel saddle coils

    Energy Technology Data Exchange (ETDEWEB)

    Rott, Michael, E-mail: Michael.Rott@ipp.mpg.de [Max Planck Institute for Plasma Physics, Boltzmannstrasse 2, D-85748 Garching (Germany); Arden, Nils; Eixenberger, Horst [Max Planck Institute for Plasma Physics, Boltzmannstrasse 2, D-85748 Garching (Germany); Klädtke, Kevin [Technical University Munich, Arcisstraße 21, D-80333 München (Germany); Teschke, Markus; Suttrop, Wolfgang [Max Planck Institute for Plasma Physics, Boltzmannstrasse 2, D-85748 Garching (Germany)

    2015-10-15

    Highlights: • Mechanical design of inverters for restricted space and good mountability. • Supply of DC link possible with thyristor converter. • Plant safety by independent overcurrent protections and a commercially arc detection system. • Ground fault detection by fast differential current measurement. • Possibilities of changing the inductance of a two windings iron core choke. - Abstract: To supply the ASDEX Upgrade's B-coils with AC current up to 1.3 kA{sub p} at 500 Hz with arbitrary waveforms, a set of 16 inverters has been designed and partially built. To keep cost and development time low, commercially available power modules are used and existing current converters feed the DC links. Three power modules are mounted in one cubicle for realizing a three level neutral point clamped (NPC) topology with lowest possible inductivity and making the most of the limited space available. The paper presents the effort and steps required to adapt standard power blocks towards the needs of the ASDEX Upgrade power supply as well as the mechanical optimizations for good mountability, flexibility and scalability. Besides, solutions for mandatory personnel safety and plant safety are presented.

  9. Electrical and mechanical adaptation of commercially available power inverter modules for BUSSARD – The power supply of ASDEX Upgrade in vessel saddle coils

    International Nuclear Information System (INIS)

    Rott, Michael; Arden, Nils; Eixenberger, Horst; Klädtke, Kevin; Teschke, Markus; Suttrop, Wolfgang

    2015-01-01

    Highlights: • Mechanical design of inverters for restricted space and good mountability. • Supply of DC link possible with thyristor converter. • Plant safety by independent overcurrent protections and a commercially arc detection system. • Ground fault detection by fast differential current measurement. • Possibilities of changing the inductance of a two windings iron core choke. - Abstract: To supply the ASDEX Upgrade's B-coils with AC current up to 1.3 kA_p at 500 Hz with arbitrary waveforms, a set of 16 inverters has been designed and partially built. To keep cost and development time low, commercially available power modules are used and existing current converters feed the DC links. Three power modules are mounted in one cubicle for realizing a three level neutral point clamped (NPC) topology with lowest possible inductivity and making the most of the limited space available. The paper presents the effort and steps required to adapt standard power blocks towards the needs of the ASDEX Upgrade power supply as well as the mechanical optimizations for good mountability, flexibility and scalability. Besides, solutions for mandatory personnel safety and plant safety are presented.

  10. A summary of the regional upgrader business plan

    International Nuclear Information System (INIS)

    Carpenter, J.S.

    1991-01-01

    The stakeholders with potential interests in an Alberta regional upgrader for bitumen, heavy oil, or refinery residue were identified and a reasonably comprehensive business plan for such an upgrader was created and analyzed. Assuming a plant accepting 30,000 bbl/d of feedstock using known hydrocracking and hydroprocessing technologies, producing a synthetic crude oil of a quality good enough to substitute for Alberta light crude in existing refineries, the risks, financing arrangements, and base case economics were examined. According to reasonable price forecasts and a socio-economic benefit assessment, there are enough benefits in such an upgrader to make its construction attractive in the near term. However, conventional investment economics make it unlikely that such a project could be supported in Alberta's current business environment. Accordingly, the business plan was revised to reduce risks and increase profitability to equity holders. The plant capacity would be doubled, with a major oil company acquiring half the output for its own production at less than half the installed cost. The rest of the plant would be owned by government and by smaller producers who would gain access to the upgrader feedstock capacity. A collar mechanism is offered to reduce bitumen price risks without impairing upgrader economics. A number of niche opportunities are also identified within the regional upgrader concept that would further enhance the return through lower costs. 2 figs., 4 tabs

  11. Use of ultrasound in petroleum residue upgradation

    Energy Technology Data Exchange (ETDEWEB)

    Sawarkar, A.N.; Pandit, A.B.; Samant, S.D.; Joshi, J.B. [Mumbai Univ., Mumbai (India). Inst. of Chemical Technology

    2009-06-15

    The importance of bottom-of-the barrel upgrading has increased in the current petroleum refining scenario because of the progressively heavier nature of crude oil. Heavy residues contain large concentrations of metals such as vanadium and nickel which foul catalysts and reduce the potential effect of residue fluidized catalytic cracking. This study showed that the cavitational energy induced by ultrasound be be successfully used to upgrade hydrocarbon mixtures. Conventional processes for the upgrading of residual feedstocks, such as thermal cracking and catalytic cracking, were carried out in the temperature range of 400-520 degrees C. Experiments were performed on 2 vacuum residues, Arabian mix vacuum residue (AMVR) and Bombay high vacuum residue (BHVR) and 1 Haldia asphalt (HA). These were subjected to acoustic cavitation for different reaction times from 15 to 120 minutes at ambient temperature and pressure. Two acoustic cavitation devices were compared, namely the ultrasonic bath and ultrasonic horn. In particular, this study compared the ability of these 2 devices to upgrade the petroleum residues to lighter, more value-added products. Different surfactants were used to examine the effect of ultrasound on upgrading the residue when emulsified in water. In order to better understand the reaction mechanism, a kinetic model was developed based on the constituents of the residue. The ultrasonic horn was found to be more effective in bringing about the upgrading than ultrasonic bath. The study also showed that the acoustic cavitation of the aqueous emulsified hydrocarbon mixture could reduce the asphaltenes content to a greater extent than the acoustic cavitation of non-emulsified hydrocarbon mixture. 20 refs., 11 tabs., 17 figs.

  12. Fitting partially upgraded oils into pipelines and refinery markets

    International Nuclear Information System (INIS)

    Flaherty, G.

    2000-01-01

    The logistics of transporting partially upgraded crudes in feeder and trunk pipeline systems is discussed. Logistic alternatives are evaluated against economic drivers for partial upgrading, and the impact of crude transportation logistics on the quality of crude that reaches refinery gates is assessed. The potential advantages of partial upgrading in the field are reviewed (including reduction of diluent required to meet pipeline density and viscosity specifications, cost and availability of diluent, limitations in diluent transportation infrastructure, increased chemical stability, increased attractiveness to refineries, shortage of refinery coking capacity, higher market value). The pros and cons of various upgrading processes, and the implications of each for producers and refiners are explained. The advantages of moving to large common streams, as opposed to the concept of 'boutique' crudes, are stressed as the surest way for producers to realize the maximum value of partially upgraded crudes

  13. Upgrade of internal events PSA model using the AESJ level-1 PSA standard for operating state

    International Nuclear Information System (INIS)

    Sato, Teruyoshi; Yoneyama, Mitsuru; Hirokawa, Naoki; Sato, Chikahiro; Sato, Eisuke; Tomizawa, Shigeatsu

    2009-01-01

    In 2003, the Atomic Energy Society of Japan (AESJ) started to develop the Level-1 Probabilistic Safety Assessment (PSA) standard of internal events for operating state (AESJ standard). The AESJ standard has been finished to be asked for public comment. Using the AESJ standard (draft version), the authors have upgraded the PSA model for Tokyo Electric Power Company (TEPCO) BWR-5 plant not only to reflect latest knowledge but also to ensure high quality of PSA model (not yet peer-reviewed) for the purpose of better operation and maintenance management of TEPCO BWR plants. For example, the categorization of structures, systems and components (SSCs) will be performed to improve nuclear reactor safety using information of risk importance. (author)

  14. FLAMMABLE GAS TECHNICAL BASIS DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    KRIPPS, L.J.

    2005-03-03

    This document describes the qualitative evaluation of frequency and consequences for DST and SST representative flammable gas accidents and associated hazardous conditions without controls. The evaluation indicated that safety-significant structures, systems and components (SSCs) and/or technical safety requirements (TSRs) were required to prevent or mitigate flammable gas accidents. Discussion on the resulting control decisions is included. This technical basis document was developed to support WP-13033, Tank Farms Documented Safety Analysis (DSA), and describes the risk binning process for the flammable gas representative accidents and associated represented hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous condition based on an evaluation of the event frequency and consequence.

  15. Energy upgrading measures improve also indoor climate

    DEFF Research Database (Denmark)

    Foldbjerg, Peter; Knudsen, Henrik Nellemose

    2014-01-01

    A new survey shows that the economy is what motivates Danish owners of single-family houses the most to start energy upgrading, and that improved indoor climate is also an important factor. After the upgrading, homeowners experience both improved economy and indoor climate. In a strategy...... to increase the number of homeowners who venture into a major energy upgrading of their house, the demonstrated positive side effects, more than energy savings, should be included in the communication to motivate homeowners. The barriers should be reduced by “taking the homeowners by the hand” and helping...... them to choose relevant energy-saving solutions as well as clarifying the financial consequences and opportunities....

  16. The basis and safety of food irradiation. Advantages of radiation treatment for food sanitation and storage

    Energy Technology Data Exchange (ETDEWEB)

    Ito, Hitoshi [Japan Atomic Energy Research Inst., Takasaki, Gunma (Japan). Takasaki Radiation Chemistry Research Establishment

    2001-09-01

    The food irradiation has the history of more than 60 years in its development. However, its commercial application has not been promoted well in Japan even though the safety of irradiated foods was confirmed. Recently, relevant authorities in 52 countries have given clearance to many commodities, and irradiated foods are commercially distributed in USA and EU countries. The international situation makes some unavoidable circumstances which can not close the commercialization of food irradiation in Japan. The present report contains the basis and application of food irradiation, and history of development in the World and Japan. Moreover, the safety of irradiated foods are demonstrated from many evidences of researches in animal feeding tests, in analysis of radiolytic products, in nutritional evaluations and in microbiological studies of irradiated foods. Especially, it makes obvious from the results of many researches that unique radiolytic products can not be produced by irradiation of foods. Because main radiation effects are induced by oxidation degradation of food components as similar to natural oxidation by heating or UV light. Radiation engineering for commercial process and identification methods of irradiated foods are also presented. (author)

  17. Canister storage building design basis accident analysis documentation

    Energy Technology Data Exchange (ETDEWEB)

    KOPELIC, S.D.

    1999-02-25

    This document provides the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

  18. New ICRP recommendations and radiation safety of an NPP

    International Nuclear Information System (INIS)

    Janzekovic, H.

    2007-01-01

    In March 2007 the fundamental radiation protection recommendations used world-widely in nuclear facilities were approved by the ICRP. Implementation of radiation safety standards in an NPP is a challenging issue related to all NPP phases from planning a site and its design to its decommissioning also because if neglected it could be very difficult if not impossible to implement improvement of radiation safety later during operation or decommissioning without a substantial cost. The standards are changing with a period of 15 years which is small regarding a prolonged lifetime of many NPPs and also foreseen lifetime of new NPPs, i.e. 60 years. The new recommendations are actually an upgrading of the ICRP 60. Among other changes new sets of wR and wT are given, as well as an update of around 50 different values related to doses. Two new concepts are also tackled i.e. terrorist attacks and protection of the environment. The influence of the new recommendations on the radiation safety of NPPs can be analysed by a selection of four renewed or new concepts: types of exposure situation, dose constraints, source-related approach and safety and security. Their implementation could lead to upgrading the radiation safety of future or existing NPPs as well as of decommissioning processes. Some of the concepts were already extensively and successfully used by designers of modifications or of new NPPs, as well as by operators. (author)

  19. Reactor safety instrumentation of Paks NPP (experience and perspective)

    International Nuclear Information System (INIS)

    Elo, S.; Hamar, K.

    1993-01-01

    The majority of the existing control and protection systems in nuclear power plants use old analog technology and design philosophy. Maintenance and the procurement of spare parts is becoming increasingly difficult. In general there is an age degradation concern. Aging degradation in nuclear power plants must be effectively managed to avoid a loss of vital safety function, shutdown of the station, a reduced power generation, or any failure leading to expensive repair. Even with the best efforts in developing reliable and long life instrumentation and control systems for nuclear power plants it is expected that these systems for most plants will require replacements during the life of the plants. The instrumentation and control system of the nuclear power plants designed during the 70's and constructed in the 80's went out-of-date since nuclear safety is not a static concept and the digital computer technology has undergone rapid improvements during the 70's and 80's. Simultaneously the operation and the maintenance of the I ampersand C system of those plants described above becomes more and more difficult and expensive. In this context the pure quality of the former Soviet designed process instrumentation system increases the needs of upgrading this system. The author reviews the main design characteristics of the reactor safety instrumentation of the Paks NPP. Further he attempts to convey the perspective on upgrading the reactor safety instrumentation as seen by the HAEC and its Nuclear Safety Inspectorate

  20. Radiation protection and safety infrastructures in Albania

    International Nuclear Information System (INIS)

    Paci, Rustem; Ylli, Fatos

    2008-01-01

    The paper intends to present the evolution and actual situation of radiation protection and safety infrastructure in Albania, focusing in its establishing and functioning in accordance with BBS and other important documents of specialized international organizations. There are described the legal framework of radiation safety, the regulatory authority, the services as well the practice of their functioning. The issue of the establishing and functioning of the radiation safety infrastructure in Albania was considered as a prerequisite for a good practices development in the peaceful uses of radiation sources . The existence of the adequate legislation and the regulatory authority, functioning based in the Basic Safety Standards (BSS), are the necessary condition providing the fulfilment of the most important issues in the mentioned field. The first document on radiation protection in Albania stated that 'for the safe use of radiation sources it is mandatory that the legal person should have a valid permission issued by Radiation Protection Commission'. A special organ was established in the Ministry of Health to supervise providing of the radiation protection measures. This organization of radiation protection showed many lacks as result of the low efficiency . The personnel monitoring, import, transport, waste management and training of workers were in charge of Institute of Nuclear Physics (INP). In 1992 an IAEA RAPAT mission visited Albania and proposed some recommendations for radiation protection improvements. The mission concluded that 'the legislation of the radiation protection should be developed'. In 1995 Albania was involved in the IAEA Model Project 'Upgrading of Radiation Protection Infrastructure'. This project, which is still in course, intended to establish the modern radiation safety infrastructures in the countries with low efficiency ones and to update and upgrade all aspects related with radiation safety: legislation and regulations, regulatory

  1. Functional Upgrading and Value Capture of Multinational Subsidiaries

    DEFF Research Database (Denmark)

    Burger, Anže; Jindra, Björn; Marek, Philipp

    2018-01-01

    survey-based business function indicators with longitudinal accounting data for a representative sample of multinational subsidiaries located in six Central and Eastern European countries (CEECs), we assess the impact of functional upgrading on foreign subsidiaries' value capture. The results provide......This paper investigates the relationship between the value capture of multinational subsidiaries and functional upgrading, which is defined as a diversification of employment from primary business functions to higher value adding activities such as ICT, R&D, marketing or logistics. By combining...... robust evidence that the breadth as well as the scope of functional upgrading induces an upward shift of subsidiaries' value added. The effect of functional upgrading is stronger in the earlier phases after entry of the foreign investor, while the long-term growth trend remains unaffected....

  2. Sizewell B nuclear power station: the basis for the decision by the Health and Safety Executive to grant consent to load fuel into the reactor

    International Nuclear Information System (INIS)

    1994-01-01

    The licensing and consent process and the basis for granting a consent for Nuclear Electric to load fuel into the Sizewell B reactor in the United Kingdom are explained. Consent was granted by the UK Nuclear Installations Inspectorate on behalf of the Health and Safety Executive on satisfactory completion of construction and those commissioning stages needed to proceed safely, and the production of a satisfactory safety case. A summary of the assessment of the safety case is appended. It covers the reactor core, coolant system structural integrity, engineered safety features, main and essential electrical system, control and instrumentation, radioactive waste management, radiological protection, fuel storage and handling, civil works and structures, fault analysis, human factors, hazard analysis, quality assurance, and decommissioning. (UK)

  3. Seismic evaluation and upgrading design of overhead roads between reactor buildings of WWER-1000 MW type NPP

    International Nuclear Information System (INIS)

    Jordanov, M.J.; Stoyanov, G.S.; Geshanov, I.H.; Kirilov, K.P.; Schuetz, W.

    2003-01-01

    This paper presents results obtained during the study of overhead roads between Reactor Building (RB) of WWER-1000 MW NPP and possible measures for their seismic upgrade. The main objective of this project is to evaluate the behavior of overhead roads under site-specific seismic loading and to determine whether this structure satisfies current international safety regulations, followed by development of upgrading concepts. Overhead roads are pre-cast RC structure, which can be divided to separate substructures. They comprise of pedestrian gallery and pipeline box, connecting reactor buildings with auxiliary building. They are mounted at approximately 10 m above ground level. The overhead roads are evaluated for Review Level Earthquake (RLE) as seismic category II structures. As seismic input motion is RLE, free field response spectra anchored to 0.2 g PGA are used with 0.5 scaling factor. Soil-Structure Interaction effects are taken into account through equivalent soil springs with frequency adjusted stiffness. In order to meet the objective of the project a technical design specification is developed for conformance with International, US and Bulgarian standards and codes, taking into account site specific conditions. The general approach is consistent with up-to-date practice for evaluation and upgrade of nuclear power plant facilities. The separate steps comprising the overall fulfillment of project's major objectives may be summarized as follows: study of all available data for initial design and as built conditions, creation of 3-D detailed finite element models for as-built structure, determination of dynamic characteristics, evaluation of adequacy of initial design under new seismic loading (calculation of D/C ratios for structural members and connections, evaluation of embedment lengths for embedded parts and rebars, deformation evaluation, stability checks), development of upgrading concepts for enhancement, verification of capability of upgraded structure

  4. Reliability Analysis of Timber Structures through NDT Data Upgrading

    DEFF Research Database (Denmark)

    Sousa, Hélder; Sørensen, John Dalsgaard; Kirkegaard, Poul Henning

    The first part of this document presents, in chapter 2, a description of timber characteristics and common used NDT and MDT for timber elements. Stochastic models for timber properties and damage accumulation models are also referred. According to timber’s properties a framework is proposed...... for a safety reassessment procedure. For that purpose a theoretical background for structural reliability assessment including probabilistic concepts for structural systems and stochastic models are given in chapter 3. System models, both series and parallel systems, are presented as well as methods...... for robustness are dealt in chapter 5. The second part of this document begins in chapter 6, where a practical application of the premise definitions and methodologies is given through the implementation of upgraded models with NDT and MDT data. Structural life-cycle is, therefore, assessed and reliability...

  5. Kongiganak Wind Turbine Replacement and System Upgrade Project

    Energy Technology Data Exchange (ETDEWEB)

    Boonstra, Patrick [Intelligent Energy Systems, Anchorage, AK (United States)

    2016-12-13

    The Native Village of Kongiganak, Alaska was awarded a grant to upgrade the braking systems on five wind turbines and upgrade the monitoring and data collection unit to insure that enough energy is available to power the utility. The project manager for this award is Intelligent Energy Systems, LLC located in Anchorage, Alaska. In addition to accomplishing these upgrades, it was the intent for a local wind tech crew to be trained in Kongiganak so that routine maintenance and future repairs will be made by local workers.

  6. Results of the safety analyses for the Greifswald and Stendal WWER nuclear power plants

    International Nuclear Information System (INIS)

    Milhem, J.L.

    1993-03-01

    Following a brief introduction of the design features of the three types of the WWER reactors, the paper deals with the main issues of the safety-related design and the most important recommendations which have been derived for upgrading measures. Furthermore some operational safety aspects of the VVER-1000 will be discussed in some detail

  7. Nuclear Safety Bureau: safety objectives and principles for the proposed ANSTO reactor

    International Nuclear Information System (INIS)

    Westall, D.

    1993-01-01

    Siting criteria and safety assessment principles were previously promulgated by the Australian Atomic Energy Commission (AAEC), and have been applied by ANSTO and the Nuclear Safety Bureau (NSB). The NSB is revising these criteria and principles to take account of evolving nuclear safety standards and practices. The NSB Safety and Siting Assessment Principles (SSAP) are presented and it is estimated that it will provide a comprehensive basis for the safety assessment of research reactors in Australia, and be applicable to all stages of a reactor project: siting: design and construction; operation; modification; and decommissioning. The SSAP are similar to the principles promulgated by the AAEC, in that probabilistic safety criteria are set for assessment of design, however these criteria are complimentary to a deterministic design basis approach. This is a similar approach to that recently published by the UK Nuclear Installations Inspectorate 4 . Siting principles are now also included, where they were previously separate, and require a consideration of the consequences of severe accidents which are an extension of accidents catered for by the design of the plant. Criteria for radiation doses due to normal operations and design basis accidents are included in the principles for safety assessment. 9 refs

  8. Advantages of viscodampers for NPPs upgrading

    International Nuclear Information System (INIS)

    Reinsch, K.H.; Schwahn, K.J.; Podrouzek, J.

    1993-01-01

    Viscodampers GERB are the ideal components for seismic upgrading of flexible pipework systems of the existing operating NPPs, of WWER-440 type, namely in primary reactor part. There are cheap, maintenance-free, simple and their installation minimizes the support forces to the structures. The quoted attests are enclosed in this paper. The references for application on Bohunice NPP upgrading are positive, and they are used in Mochovce and Temelin NPP as well

  9. Radiation and waste safety: Strengthening national capabilities

    International Nuclear Information System (INIS)

    Barretto, P.; Webb, G.; Mrabit, K.

    1997-01-01

    For many years, the IAEA has been collecting information on national infrastructures for assuring safety in applications of nuclear and radiation technologies. For more than a decade, from 1984-95, information relevant to radiation safety particularly was obtained through more than 60 expert missions undertaken by Radiation Protection Advisory Teams (RAPATs) and follow-up technical visits and expert missions. The RAPAT programme documented major weaknesses and the reports provided useful background for preparation of national requests for IAEA technical assistance. Building on this experience and subsequent policy reviews, the IAEA took steps to more systematically evaluate the needs for technical assistance in areas of nuclear and radiation safety. The outcome was the development of an integrated system designed to more closely assess national priorities and needs for upgrading their infrastructures for radiation and waste safety

  10. RBMK nuclear reactors: Proposals for instrumentation and control improvements to enhanced safety and availability. IEC technical report of type 3. Working material

    International Nuclear Information System (INIS)

    1995-01-01

    The present material presents a CD+V draft report ''RBMK nuclear reactors: Proposals for instrumentation and control improvements to enhance safety and availability'' prepared by the Joint IEC/IAEA team during 1993-1995. Experience has demonstrated the need to improve the safety instrumentation of the RBMK type reactors using well proven modern technology. The working group identified the upgrades and changes of the highest priority based on the evaluation of the RBMK systems and the events where the instrumentation was found to be inadequate for safe operation. The subjects discussed in this document were not selected on a systematic basis but were selected by the IEC and IAEA experts as considered to be appropriate to the activities of the IEC and for which technical experience was available. The items identified therefore do not reflect any ranking of the safety issues or any priority or impact on safety of any of the measures were they to be implemented. Many important safety issued and areas where physical measures are required to improve safety have been omitted and indeed not even acknowledged in this document. The recommendations presented in the document differ from those normally produced by the IEC in the form of standards as they are of a transitory nature and some have already been overtaken by the continuing process of improvements to plant safety. Figs and tabs

  11. Academic Training - LHC luminosity upgrade: detector challenges

    CERN Multimedia

    Françoise Benz

    2006-01-01

    ACADEMIC TRAINING LECTURE SERIES 13, 14, 15, March, from 11:00 to 12:00 - 16 March from 10:00 to 12:00 Main Auditorium, bldg. 500 on 14, 15 March, Council Room on 13, 16 March LHC luminosity upgrade: detector challenges A. De Roeck / CERN-PH, D. Bortoletto / Purdue Univ. USA, R. Wigmans / Texas, Tech Univ. USA, W. Riegler / CERN-PH, W. Smith / Wisconsin Univ. USA The upgrade of the LHC machine towards higher luminosity (1035 cm-2s-1) has been studied over the last few years. These studies have investigated scenarios to achieve the increase in peak luminosity by an order of magnitude, as well as the physics potential of such an upgrade and the impact of a machine upgrade on the LHC DETECTORS. This series of lectures will cover the following topics: Physics motivation and machine scenarios for an order of magnitude increase in the LHC peak luminosity (lecture 1) Detector challenges including overview of ideas for R&D programs by the LHC experiments: tracking and calorimetry, other new detector ...

  12. Upgrade of the ATLAS Tile Calorimeter Electronics

    CERN Document Server

    Moreno, P; The ATLAS collaboration

    2014-01-01

    The Tile Calorimeter (TileCal) is the hadronic calorimeter covering the central region of the ATLAS experiment at LHC. The TileCal readout consists of about 10000 channels. The bulk of its upgrade will occur for the High Luminosity LHC phase (phase 2) where the peak luminosity will increase 5x compared to the design luminosity (10^34 cm−2s−1) but with maintained energy (i.e. 7+7 TeV). An additional increase of the average luminosity with a factor of 2 can be achieved by luminosity leveling. This upgrade is expected to happen around 2023. The TileCal upgrade aims at replacing the majority of the on- and off-detector electronics to the extent that all calorimeter signals will be digitized and sent to the off-detector electronics in the counting room. To achieve the required reliability, redundancy has been introduced at different levels. Three different options are presently being investigated for the front-end electronic upgrade. Extensive test beam studies will determine which option will be selected. 10 ...

  13. Upgrade of the CMS Global Muon Trigger

    CERN Document Server

    Lingemann, Joschka; Sakulin, Hannes; Jeitler, Manfred; Stahl, Achim

    2015-01-01

    The increase in center-of-mass energy and luminosity for Run 2 of the Large Hadron Collider pose new challenges for the trigger systems of the experiments. To keep triggering with a similar performance as in Run 1, the CMS muon trigger is currently being upgraded. The new algorithms will provide higher resolution, especially for the muon transverse momentum and will make use of isolation criteria that combine calorimeter with muon information already in the level-1 trigger. The demands of the new algorithms can only be met by upgrading the level-1 trigger system to new powerful FPGAs with high bandwidth I/O. The processing boards will be based on the new microTCA standard. We report on the planned algorithms for the upgraded Global Muon Trigger (GMT) which combines information from the muon trigger sub-systems and assigns the isolation variable. The upgraded GMT will be implemented using a Master Processor 7 card, built by Imperial College, that features a large Xilinx Virtex 7 FPGA. Up to 72 optical links at...

  14. LHCb VELO Upgrade

    CERN Document Server

    van Beuzekom, Martin; Ketel, Tjeerd; Gershon, Timothy; Parkes, Christopher; Reid, Matthew

    2011-01-01

    The VErtex LOcator (VELO) is a vital piece of apparatus for allowing precision measurements in hadronic physics. It provides not only superb impact parameter resolutions but also excellent momentum resolution, both important discriminating tools for precision high energy physics. This poster focuses on the R&D going into the future LHCb VELO detector. At present there are two proposed options for the upgrade; pixel chips or strip detectors. The LHCb upgrade is designed with higher luminosities and increased yields in mind. In order to get more out of the LHCb detector changes to the front end electronics will have to be made. At present, the first level hardware trigger is sets a limiting factor on the maximum efficiency for hadronic channels. As the VELO is positioned so close the proton-proton interaction region, whatever the choice of sensor, we will require efficient cooling and some proposed solutions are outlined. The LHCb TimePix telescope has had a very successful years running, with various devic...

  15. The JLab 12 GeV Energy Upgrade of CEBAF for QCD and Hadronic Physics

    International Nuclear Information System (INIS)

    Lawrence Cardman; Leigh Harwood

    2007-01-01

    CEBAF at Jefferson Lab is a 5-pass, recirculating cw electron linac operating at ∼6 GeV and devoted to basic research in nuclear physics. The 12 GeV Upgrade is a major project, sponsored by the DOE Office of Nuclear Physics, that will expand its research capabilities substantially by doubling the maximum energy and adding major new experimental apparatus. We anticipate that the project will receive Critical Decision 2 approval this year and begin construction in 2008. The research program motivating the Upgrade includes: the study of hybrid mesons, which involve excited states of the glue, to explore the nature of quark confinement; dramatic improvements in our understanding of the QCD structure of the hadrons through the extension of our knowledge of their parton distribution functions to high xBjorken, where they are dominated by underlying valence quark structure, and a program of nucleon ''tomography'' via measurements of the Generalized Parton Distributions (GPDs), a broad program of experiments in the physics of nuclei that aims to understand the QCD basis for the nucleon-nucleon force and how nucleons and mesons arise as an approximation to the underlying quark-gluon structure; and precision tests of the Standard Model through parity violating deep inelastic and Moeller scattering. The Upgrade includes: doubling the accelerating voltages of the linacs by adding 10 new high-performance cryomodules; the requisite expansion of the 2K cryogenics plant and rf power systems to support these cryomodules; upgrading the beam transport system from 6 to 12 GeV through extensive re-use and/or modification of existing hardware; and the addition of one recirculation arc, a new experimental area, and the beamline to it; and the construction of major new experimental equipment for the GPD, high-xBjorken, and hybrid meson programs. The presentation will describe the science briefly and provide some details about the accelerator plans

  16. Industrial Clusters and Social Upgrading in Developing Countries

    DEFF Research Database (Denmark)

    Pyke, Frank; Lund-Thomsen, Peter

    In this article, we explore the relationship between industrial clusters and social upgrading in developing countries. Our article focuses on the hitherto little-considered influence of the economic and regulatory environment on the social upgrading of a cluster and on its governance system....... In doing so, we develop an analytical framework that seeks to explain how the enabling environment and different actors in cluster governance can either facilitate and/or hinder the process of social upgrading in cluster settings in developing countries. Finally, the conclusion outlines our main findings...

  17. LHCb DAQ network upgrade tests

    CERN Document Server

    Pisani, Flavio

    2013-01-01

    My project concerned the evaluation of new technologies for the DAQ network upgrade of LHCb. The first part consisted in developing and Open Flow-based Clos network. This new technology is very interesting and powerful but, as shown by the results, it still needs further improvements. The second part consisted in testing and benchmarking 40GbE network equipment: Mellanox MT27500, Chelsio T580 and Huawei Cloud Engine 12804. An event-building simulation is currently been performed in order to check the feasibility of the DAQ network upgrade in LS2. The first results are promising.

  18. Status of the University of Missouri-Columbia Research Reactor upgrade

    International Nuclear Information System (INIS)

    McKibben, J.C.; Edwards, C.B. Jr.; Meyer, W.A. Jr.; Kim, S.S.

    1990-01-01

    The University of Missouri-Columbia (MU) Research Reactor Facility staff is in the process of upgrading the operational and research capabilities of the reactor and associated facilities. The upgrades include an extended life aluminide fuel element, a power increase, improved instrumentation and control equipment, a cold neutron source, a building addition, and improved research instrumentation and equipment. These upgrades will greatly enhance the capabilities of the facility and the research programs. This paper discusses the parts of the upgrade and current status of implementation. (author)

  19. Status of the University of Missouri-Columbia Research Reactor upgrade

    Energy Technology Data Exchange (ETDEWEB)

    McKibben, J C; Edwards, Jr, C B; Meyer, Jr, W A [MU Research Reactor, Columbia, MO (United States); Kim, S S [Idaho Nuclear Engineering Laboratory, Idaho Falls, ID (United States)

    1990-05-01

    The University of Missouri-Columbia (MU) Research Reactor Facility staff is in the process of upgrading the operational and research capabilities of the reactor and associated facilities. The upgrades include an extended life aluminide fuel element, a power increase, improved instrumentation and control equipment, a cold neutron source, a building addition, and improved research instrumentation and equipment. These upgrades will greatly enhance the capabilities of the facility and the research programs. This paper discusses the parts of the upgrade and current status of implementation. (author)

  20. Application of Microprocessor-Based Equipment in Nuclear Power Plants - Technical Basis for a Qualification Methodology

    International Nuclear Information System (INIS)

    Korsah, K.

    2001-01-01

    This document (1) summarizes the most significant findings of the ''Qualification of Advanced Instrumentation and Control (I and C) Systems'' program initiated by the Nuclear Regulatory Commission (NRC); (2) documents a comparative analysis of U.S. and European qualification standards; and (3) provides recommendations for enhancing regulatory guidance for environmental qualification of microprocessor-based safety-related systems. Safety-related I and C system upgrades of present-day nuclear power plants, as well as I and C systems of Advanced Light-Water Reactors (ALWRs), are expected to make increasing use of microprocessor-based technology. The Nuclear Regulatory Commission (NRC) recognized that the use of such technology may pose environmental qualification challenges different from current, analog-based I and C systems. Hence, it initiated the ''Qualification of Advanced Instrumentation and Control Systems'' program. The objectives of this confirmatory research project are to (1) identify any unique environmental-stress-related failure modes posed by digital technologies and their potential impact on the safety systems and (2) develop the technical basis for regulatory guidance using these findings. Previous findings from this study have been documented in several technical reports. This final report in the series documents a comparative analysis of two environmental qualification standards--Institute of Electrical and Electronics Engineers (IEEE) Std 323-1983 and International Electrotechnical Commission (IEC) 60780 (1998)--and provides recommendations for environmental qualification of microprocessor-based systems based on this analysis as well as on the findings documented in the previous reports. The two standards were chosen for this analysis because IEEE 323 is the standard used in the U.S. for the qualification of safety-related equipment in nuclear power plants, and IEC 60780 is its European counterpart. In addition, the IEC document was published in