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Sample records for westinghouse ap1000 advanced

  1. The Westinghouse Advanced Passive Pressurized Water Reactor, AP1000

    International Nuclear Information System (INIS)

    Schene, R.

    2009-01-01

    Featuring proven technology and innovative passive safety systems, the Westinghouse AP1000 pressurized water reactor can achieve competitive generation costs in the current electricity market without emitting harmful greenhouse gases and further harming the environment. Westinghouse Electric Company, the pioneer in nuclear energy once again sets a new industry standard with the AP1000. The AP1000 is a two-loop pressurized water reactor that uses simplified, innovative and effective approach to safety. With a gross power rating of 3415 megawatt thermal and a nominal net electrical output of 1117 megawatt electric, the AP1000 is ideal for new base load generation. The AP1000 is the safest and most economical nuclear power plant available in the worldwide commercial marketplace, and is the only Generation III+ reactor to receive a design certification from the U.S. Nuclear Regulatory Commission (NRC). Based on nearly 20 years of research and development, the AP1000 builds and improves upon the established technology of major components used in current Westinghouse designed plants. These components, including steam generators, digital instrumentation and controls, fuel, pressurizers, and reactor vessels, are currently in use around the world and have years of proven, reliable operating experience. Historically, Westinghouse plant designs and technology have forged the cutting edge technology of nuclear plant around the world. Today, nearly 50 percent of the world's 440 nuclear plants are based on Westinghouse technology. Westinghouse continues to be the nuclear industry's global leader. (author)

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  3. Westinghouse AP1000 licensing maturity

    International Nuclear Information System (INIS)

    Schulz, T.; Vijuk, R.P.

    2005-01-01

    The Westinghouse AP1000 Program is aimed at making available a nuclear power plant that is economical in the U.S deregulated electrical power industry in the near-term. The AP1000 is two-loop 1000 MWe pressurizer water reactor (PWR). It is an up rated version of the AP600. The AP1000 uses passive safety systems to provide significant and measurable improvements in plant simplification, safety, reliability, investment protection and plant costs. The AP1000 uses proven technology, which builds on over 35 years of operating PWR experience. The AP1000 received Final Design Approval by the United States Nuclear Regulatory Commission (U.S. NRC) in September 2004. The AP1000 meets the US utility requirements. The AP1000 and its sister plant the AP600 have gone through a very through and complete licensing review. This paper describes the U.S. NRC review efforts of both the AP600 and the AP1000. The detail of the review and the independent calculations, evaluations and testing is discussed. The AP600 licensing documentation was submitted in 1992. The U.S. NRC granted Final Design Approval in 1999. During the intervening 7 years, the U.S. NRC asked thousands of questions, performed independent safety analysis, audited Westinghouse calculations and analysis, and performed independent testing. The more significant areas of discussion will be described. For the AP1000 Westinghouse first engaged the U.S. NRC in pre-certification discussions to define the extent of the review required, since the design is so similar to the AP600. The AP1000 licensing documentation was submitted in March 2002. The U.S. NRC granted Final Design Approval in September 2004. During the intervening 2 1/2 years, the U.S. NRC asked hundreds of questions, performed independent safety analysis, audited Westinghouse calculations and analysis, and performed independent testing. The more significant areas of discussion will be described. The implications of this review and approval on AP1000 applications in

  4. Westinghouse AP 1000 program status

    International Nuclear Information System (INIS)

    Doehnert, B.

    2002-01-01

    The project 1000 is presented and features are discussed in the paper. Design maturity is characterized by 1300 man-year / $400 million design and testing effort, more than 12 000 design documents completed; 3D computer model developed. It includes structures, equipment, small / large pipe, cable trays, ducts etc. Licensing Maturity is determined by a very thorough and complete NRC review of AP600; 110 man-year effort (NRC) over 6 years, $30 million; independent, confirmatory plant analysis; independent, confirmatory plant testing (ROSA, OSU); over 7400 questions answered, no open items; over 380 meeting with NRC, 43 meetings with ACRS. NRC Design Certification is issued in December 1999. Reasons for developing AP 1000 and design changes are presented. Economic analysis shows an expectation for payback within 20 years. AP1000 provides 75% power uprate for 15% increment in capital cost. AP1000 meets new plant economic targets in the near term

  5. Westinghouse plans global new builds for AP1000

    Energy Technology Data Exchange (ETDEWEB)

    Mitev, Lubomir [NucNet, Brussels (Belgium)

    2014-10-15

    Interview with Danny Roderick, Westinghouse Electric Company, President and Chief Executive Officer since September 2012, about perspectives and future plans for AP1000 new build worldwide. Within three to four years there wille be 'shovels in the ground' for three new AP1000 reactors in the UK, as well as new units in China and Bulgaria. Four AP1000 reactors are under construction in the United States at Vogtle and VC Summer, and soon at Turkey Point. Additionally Danny Roderick spoke about the acquisition of NuGen, technology transfer, the influence of the Ukraine crises on the nuclear market in East Europe and the future need for more nuclear worldwide and in the UK and Bulgaria.

  6. Westinghouse plans global new builds for AP1000

    International Nuclear Information System (INIS)

    Mitev, Lubomir

    2014-01-01

    Interview with Danny Roderick, Westinghouse Electric Company, President and Chief Executive Officer since September 2012, about perspectives and future plans for AP1000 new build worldwide. Within three to four years there wille be 'shovels in the ground' for three new AP1000 reactors in the UK, as well as new units in China and Bulgaria. Four AP1000 reactors are under construction in the United States at Vogtle and VC Summer, and soon at Turkey Point. Additionally Danny Roderick spoke about the acquisition of NuGen, technology transfer, the influence of the Ukraine crises on the nuclear market in East Europe and the future need for more nuclear worldwide and in the UK and Bulgaria.

  7. Westinghouse AP1000 Electrical Generation Costs - Meeting Marketplace Requirements

    International Nuclear Information System (INIS)

    Paulson, C. Keith

    2002-01-01

    completion. These plants are the Westinghouse advanced passive designs - AP600 and AP1000 - both of which have verifiable engineering design packages that are more than 50 percent complete. (author)

  8. Westinghouse and nuclear renaissance. The Westinghouse AP1000 - a technology solution for Slovakia

    International Nuclear Information System (INIS)

    Kirst, M.

    2009-01-01

    The Westinghouse AP1000 nuclear reactor design has been chosen by both China and the United States as the preferred technology in their new reactor programs. With four reactors in China and six in the United States under contract, in addition to the only Generation III+ design with NRC certification as well as the European Utility Requirements certification, the AP1000 has both a strong global customer base and regulatory certainty to facilitate its adoption in the Slovak Republic. (author)

  9. AP1000, a nuclear central of advanced design; AP1000, una central nuclear de diseno avanzado

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez M, N.; Viais J, J. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: nhm@nuclear.inin.mx

    2005-07-01

    The AP1000 is a design of a nuclear reactor of pressurized water (PWR) of 1000 M We with characteristic of safety in a passive way; besides presenting simplifications in the systems of the plant, the construction, the maintenance and the safety, the AP1000 is a design that uses technology endorsed by those but of 30 years of operational experience of the PWR reactors. The program AP1000 of Westinghouse is focused to the implementation of the plant to provide improvements in the economy of the same one and it is a design that is derived directly of the AP600 designs. On September 13, 2004 the US-NRC (for their initials in United States- Nuclear Regulatory Commission) approved the final design of the AP1000, now Westinghouse and the US-NRC are working on the whole in a complete program for the certification. (Author)

  10. Essence and characteristics of the Westinghouse technology AP1000

    International Nuclear Information System (INIS)

    Llovet, Ricardo

    2014-01-01

    The AP1000 nuclear power plant can place the reactor in a Safe Shutdown Condition within the first 72 hours of a Station Blackout, without the use of AC power or operator action •With some operator action after 3 days, the AP1000 nuclear power plant continues to maintain reactor core cooling and Spent Fuel Pool cooling indefinitely •The AP1000 nuclear power plant has superior coping capabilities as well as significantly reduced risk for core damage

  11. Learning through delivery, Westinghouse AP1000 plant construction

    International Nuclear Information System (INIS)

    Gorgemans, J.; Hinman, R.D.; Steuck, C.M.; Greco, P.L.

    2014-01-01

    The AP1000 plant, which is a 1100 MWe class pressurized water reactor with passive safety features, is designed around a conventional 2 loop, 2 steam generator primary system configuration with 2 hot legs, 4 reactor coolant pumps directly mounted in the steam generator lower head and 4 cold legs. A particular feature of AP1000 is its modular construction to minimize the time and cost of construction. Modular construction allows activities to be run in parallel, it allows more activities to be performed in a controlled factory instead of in the field, and it provides a better level of quality. The AP1000 plant design includes 106 structural modules and 52 mechanical modules. Structural modules include all penetrations for piping, cable trays, HVAC duct runs, and all reinforcement for pipe, equipment hangers, and supports. Structural modules are shipped in sub-modules to support transportation by rail or truck or barge. Mechanical modules contain equipment such as pumps, tanks, heat exchangers, air-handling units, and filters along with interconnecting pipes, valves, instruments, wiring and support services. Modular construction requires strong coordination between engineering, supply chain and construction. A total of 8 AP1000 units are currently under construction in China and in the United States. The lessons learned and best practices of each new AP1000 construction are systematically incorporated into the standard design. (A.C.)

  12. AP1000, a nuclear central of advanced design

    International Nuclear Information System (INIS)

    Hernandez M, N.; Viais J, J.

    2005-01-01

    The AP1000 is a design of a nuclear reactor of pressurized water (PWR) of 1000 M We with characteristic of safety in a passive way; besides presenting simplifications in the systems of the plant, the construction, the maintenance and the safety, the AP1000 is a design that uses technology endorsed by those but of 30 years of operational experience of the PWR reactors. The program AP1000 of Westinghouse is focused to the implementation of the plant to provide improvements in the economy of the same one and it is a design that is derived directly of the AP600 designs. On September 13, 2004 the US-NRC (for their initials in United States- Nuclear Regulatory Commission) approved the final design of the AP1000, now Westinghouse and the US-NRC are working on the whole in a complete program for the certification. (Author)

  13. Westinghouse AP1000® PWR: Meeting Customer Commitments and Market Needs

    International Nuclear Information System (INIS)

    Shulyak, Nick

    2014-01-01

    Westinghouse Electric Company once again sets a new industry standard with the AP1000 reactor. Historically, Westinghouse plant designs and technology have forged the cutting edge of worldwide nuclear technology. Today, about 50 percent of the world's 440 nuclear plants are based on Westinghouse technology. The AP1000 is the safest and most economical nuclear power plant available in the worldwide commercial marketplace, and is the only Generation III+ reactor to receive Design Certification from the U.S. Nuclear Regulatory Commission (NRC). The AP1000 features proven technology, innovative passive safety systems and offers: Unequalled safety, Economic competitiveness, Improved and more efficient operations. The AP1000 builds and improves upon the established technology of major components used in current Westinghouse-designed plants with proven, reliable operating experience over the past 50 years. These components include: Steam generators, Digital instrumentation and controls, Fuel, Pressurizers, Reactor vessels. Simplification was a major design objective for the AP1000. The simplified plant design includes overall safely systems, normal operating systems, the control room, construction techniques, and instrumentation and control systems. The result is a plant that is easier and less expensive to build, operate and maintain. The AP1000 design saves money and time with an accelerated construction time period of approximately 36 months, from the pouring of first concrete to the loading of fuel. Also, the innovative AP1000 features: 50% fewer safety-related valves, 80% less safety-related piping, 85% less control cable, 35% fewer pumps , 45% less seismic building volume. Eight AP1000 units under construction worldwide-Four units in China-Four units in the United States. (author)

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

    International Nuclear Information System (INIS)

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

    2015-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  16. 76 FR 73720 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Westinghouse AP1000...

    Science.gov (United States)

    2011-11-29

    ... NUCLEAR REGULATORY COMMISSION [NRC-2011-0272] Knowledge and Abilities Catalog for Nuclear Power...) is issuing for public comment a draft NUREG, NUREG-2103, Revision 0, ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Westinghouse AP1000 Pressurized-Water Reactors. DATES: Submit...

  17. Current status of generation III nuclear power and assessment of AP1000 developed by Westinghouse

    International Nuclear Information System (INIS)

    Zhang Mingchang

    2005-01-01

    In order to make greater contributions to the environment, new nuclear power systems will be needed to meet the increase of electricity demand and to replace plants to be decommissioned. A series of new designs, so called Generation III and Generation III +, are being developed to ensure their deployment in a Near-Term Deployment Road-map in US by 2010 and in Europe by 2015. The AP1000, developed by Westinghouse, is a two-loop 1000 MWe PWR with passive safety features and extensive simplifications to enhance its competitiveness in cost and tariff. It is the first Generation III + plant receiving the Final Design Approval by the US NRC. This paper briefly describes AP1000 design features and technical specifications, and presents a more detailed design evaluation with reference to relevant literatures. Both the opportunity and challenges for nuclear power development in China during the first decade of the 21 st century in a historic transition from Gen II to Gen III are analyzed. The key is to balance risks and benefits if the first AP1000 to be settled down in China. (author)

  18. The Westinghouse AP1000®: Passive, Proven Technology to Meet European Energy Demands

    International Nuclear Information System (INIS)

    Haspel, N.

    2015-01-01

    Even though several years ago nuclear power was merely considered to be an “optimistic future assessment”, the world-wide renaissance of nuclear power has become reality! The economical and climate-friendly nuclear power generation is internationally regarded to be in an evident upturn. The 435 nuclear power plants in operation worldwide are being modernized and the capacity is increased continuously. Furthermore, to date, 42 power plants are under construction, another 81 are already being applied for and or definitely planned. The global total net capacity out of nuclear power will increase accordingly in the upcoming years from currently 372 to more than 500 GWe, which presents an increase of more than one third. Westinghouse’s contribution hereto is considerable: At the present time, 4 power plants of the series AP1000 ® are under construction. To begin with, 2 units each are under construction at the Chinese sites Sanmen and Haiyang, another 4 per site are being planned. In the USA, Westinghouse has been contracted with a Engineering, Procurement and Construction (EPC) project for a total of 4 power plant units at the Vogtle and V.C. Summer. Also in Europe, the plans to construct new plants are meanwhile very specific and many countries have formally established the marginal conditions for new nuclear projects. The AP1000 ® , with its medium output capacity, is ideally positioned for many markets and can – as a twin unit – also cover large capacity demands. At the present time, Westinghouse, with its AP1000 ® , participates in the so-called GDA (Generic Design Assessment) process in Great Britain, where the British regulatory authorities conduct an assessment and evaluation of the safety aspects of this plant design in a defined multilevel process. The successful conclusion of this process ultimately leads to a “Design Acceptance Confirmation”, which will basically make the construction of the plant in Great Britain possible. (author)

  19. Oregon state university's advanced plant experiment (APEX) AP1000 integral facility test program

    International Nuclear Information System (INIS)

    Reyes, J.N.; Groome, J.T.; Woods, B.G.; Young, E.; Abel, K.; Wu, Q.

    2005-01-01

    Oregon State University (OSU) has recently completed a three year study of the thermal hydraulic behavior of the Westinghouse AP1000 passive safety systems. Eleven Design Basis Accident (DBA) scenarios, sponsored by the U.S. Department of Energy (DOE) with technical support from Westinghouse Electric, were simulated in OSU's Advanced Plant Experiment (APEX)-1000. The OSU test program was conducted within the purview of the requirements of 10CFR50 Appendix B, NQA-1 and 10 CFR 21 and the test data was used to provide benchmarks for computer codes used in the final design approval of the AP1000. In addition to the DOE certification testing, OSU conducted eleven confirmatory tests for the U.S. Nuclear Regulatory Commission. This paper presents the test program objectives, a description of the APEX-1000 test facility and an overview of the test matrix that was conducted in support of plant certification. (authors)

  20. Application of MSHIM core control strategy for westinghouse AP1000 nuclear power plant

    International Nuclear Information System (INIS)

    Onoue, Masaaki; Kawanishi, Tomohiro; Carlson, William R.; Morita, Toshio

    2003-01-01

    Westinghouse has developed a new core control strategy, in which two independently moving Rod Cluster Control Assembly (RCCA) groups are utilized for core control; one group for reactivity/temperature control, the other for axial power distribution (Axial Offset) control. This control procedure eliminates the need for Chemical Shim adjustments during power maneuvers, such as load follow, and is designated MSHIM (Mechanical Shim). This core control strategy is utilized in the AP1000. In the AP1000, it is possible to perform MSHIM load follow maneuvers for up to 95% of cycle life without changing the soluble boron concentration in the moderator. This core control strategy has been evaluated, via computer simulations, to provide appropriate margins to core and fuel design limits during normal operation maneuvers (including load follow operations) and also during anticipated Condition II accident transients. The primary benefits of MSHIM as a control strategy are as follows; Power change operation can be totally automated due to the elimination of boron concentration adjustments. Full load follow capability is achievable for up to more than 95% of cycle life. Load follow operations performed solely by mechanical devices results in a significant reduction in the boron system requirements and a significant reduction in daily effluent to be processed. (author)

  1. AP1000R pressurised water reactor project in china advances toward completion

    International Nuclear Information System (INIS)

    Harrop, G.

    2014-01-01

    The AP1000 R pressurised water reactor (PWR) project in China is the first deployment of its first-of-a-kind Generation III+ technology, making it one of most internationally important and industry-significant new build projects. The innovative AP1000 PWR design contains advanced passive safety and performance features that involve fewer active safety components than a traditional plant, thereby reducing the site footprint. The AP1000 reactor is the first and only Generation III+ nuclear power plant to be granted design certification by the United States Nuclear Regulatory Commission, and it has received an Interim Design Acceptance Confirmation from the Office for Nuclear Regulation and an Interim Statement of Design Acceptability from the Environment Agency in the United Kingdom. Construction and testing of dual AP1000 PWR units is currently in progress in each of two coastal sites in the People's Republic of China: Sanmen (Zhejiang Province) and Haiyang (Shandong Province). Since the initial contract award in 2007, the Westinghouse Consortium has worked in concert with the owners to construct the plants using innovative structural and mechanical modules. Uniquely designed plant components and essential instrumentation and control systems have been manufactured, delivered, and installed at the plants. Numerous personnel, including future reactor operators, have been trained at both the Sanmen and Haiyang sites, and technology transfer of technical documents and computer codes is well underway. The commercial operation dates are now nearing for Sanmen Unit 1 and Haiyang Unit 1, the first two units scheduled for completion. Consequently, these units are now in advanced stages of completion and present activities include planning and preparation for pre-operational testing, system turnover, and commissioning leading to fuel load, and eventual commercial operation. These activities are pioneering, in that they have never before been performed for a new build of

  2. Reactor coolant pump type RUV for Westinghouse Electric Company LLC reactor AP1000 TM

    International Nuclear Information System (INIS)

    Baumgarten, S.; Brecht, B.; Bruhns, U.; Fehring, P.

    2010-01-01

    The RUV is a reactor coolant pump, specially designed for the Westinghouse Electric Company LLC AP1000 TM reactor. It is a hermetically sealed, wet winding motor pump. The RUV is a very compact, vertical pump/motor unit, designed to fit into the compartment next to the reactor pressure vessel. Each of the two steam generators has two pump casings welded to the channel head by the suction nozzle. The pump/motor unit consists of a pump part, where a semi-axial impeller/diffuser combination is mounted in a one-piece pump casing. Computational Fluid Dynamics methods combined with various hydraulic tests in a 1:2 scale hydraulic test assure full compliance with the specific customer requirements. A short and rigid shaft, supported by a radial bearing, connects the impeller with the high inertia flywheel. This flywheel consists of a one-piece forged stainless steel cylinder, with an option for several smaller heavy metal cylinders inside. The flywheel is located inside the thermal barrier, which forms part of the pressure boundary. A specific arrangement of cooling water circuits guarantees a homogeneous temperature distribution in and around the flywheel, minimizes the friction losses of the flywheel and protects the motor from hot coolant. The driving torque is transmitted by the motor shaft, which itself is supported by two radial bearings. A three-phase, high-voltage squirrel-cage induction motor generates the driving torque. Due to the wet winding concept it is possible to achieve positive effects regarding motor lifetime. The cooling water is forced through the stator windings and the gap between rotor and stator by an auxiliary impeller. Furthermore, this wet winding motor concept has higher efficiency as compared to a canned motor since there are no eddy current losses. As part of the design process and in addition to the hydraulic scale model, a complete half scale model pump was built. It was used to verify the calculations performed like coast

  3. AP1000. The PWR revisited

    International Nuclear Information System (INIS)

    Gaio, P.

    2006-01-01

    The distinguishing features of Westinghouse's AP1000 advanced passive pressurized water reactor are highlighted. In particular, the AP1000's passive safety features are described as well as their implications for simplifying the design, construction, and operation of this design compared to currently operating plants, and significantly increasing safety margins over current plants as well. The AP1000 design specifically incorporates the knowledge acquired from the substantial accumulation of power reactor operating experience and benefits from the application of the Probabilistic Risk Assessment in the design process itself. The AP1000 design has been certified by the US Nuclear Regulatory Commission under its new rules for licensing new nuclear plants, 10 CFR Part 52, and is the subject of six combined Construction and Operating License applications now being developed. Currently the AP1000 design is being assessed against the EUR Rev C requirements for new nuclear power plants in Europe. (author)

  4. AP1000 Design for Security

    International Nuclear Information System (INIS)

    Long, L.B.; Cummins, W.E.; Winters, J.W.

    2006-01-01

    Nuclear power plants are protected from potential security threats through a combination of robust structures around the primary system and other vital equipment, security systems and equipment, and defensive strategy. The overall objective for nuclear power plant security is to protect public health and safety by ensuring that attacks or sabotage do not challenge the ability to safely shutdown the plant or protect from radiological releases. In addition, plants have systems, features and operational strategies to cope with external conditions, such as loss of offsite power, which could be created as part of an attack. Westinghouse considered potential security threats during design of the AP1000 PWR. The differences in plant configuration, safety system design, and safe shutdown equipment between existing plants and AP1000 affect potential vulnerabilities. This paper provides an evaluation of AP1000 with respect to vulnerabilities to security threats. The AP1000 design differs from the design of operating PWRs in the US in the configuration and the functional requirements for safety systems. These differences are intentional departures from conventional PWR designs which simplify plant design and enhance overall safety. The differences between the AP1000 PWR and conventional PWRs can impact vulnerabilities to security threats. The NRC addressed security concerns as part of their reviews for AP1000 Design Certification, and did not identify any security issues of concern. However, much of the detailed security design information for the AP1000 was deferred to the combined Construction and Operating License (COL) phase as many of the security issues are site-specific. Therefore, NRC review of security issues related to the AP1000 is not necessarily complete. Further, since the AP1000 plant design differs from existing PWRs, it is not obvious that the analyses and assessments prepared for existing plants also apply to the AP1000. We conclude that, overall, the AP1000

  5. 77 FR 56241 - Notice of Withdrawal of Final Design Approval; Westinghouse Electric Company; Advanced Passive 1000

    Science.gov (United States)

    2012-09-12

    ... NUCLEAR REGULATORY COMMISSION [NRC-2010-0131] Notice of Withdrawal of Final Design Approval; Westinghouse Electric Company; Advanced Passive 1000 By letter dated December 10, 2010, Westinghouse Electric... final design approval (FDA) for the Advanced Passive 1000 (AP1000) design upon the completion of...

  6. Safety features and research needs of westinghouse advanced reactors

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  7. AP1000 construction schedule

    International Nuclear Information System (INIS)

    Winters, J.W.

    2001-01-01

    Westinghouse performed this study as part of EPRI interest in advancing the use of computer aided processes to reduce the cost of nuclear power plants. EPRI believed that if one could relate appropriate portions of an advanced light water reactor plant model to activities in its construction sequence, and this relationship could be portrayed visually, then optimization of the construction sequence could be developed as never before. By seeing a 3-D representation of the plant at any point in its construction sequence, more informed decisions can be made on the feasibility or attractiveness of follow on or parallel steps in the sequence. The 3-D representation of construction as a function of time (4-D) could also increase the confidence of potential investors concerning the viability of the schedule and the plant ultimate cost. This study performed by Westinghouse confirmed that it is useful to be able to visualize a plant construction in 3-D as a function of time in order to optimize the sequence of construction activities. (author)

  8. AP1000{sup TM} plant modularization

    Energy Technology Data Exchange (ETDEWEB)

    Cantarero L, C.; Demetri, K. J. [Westinghouse Electric Co., 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States); Quintero C, F. P., E-mail: cantarc@westinghouse.com [Westinghouse Electric Spain, Padilla 17, 28006 Madrid (Spain)

    2016-09-15

    The AP1000{sup TM} plant is an 1100 M We pressurized water reactor (PWR) with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. Modules are used extensively in the design of the AP1000 plant nuclear island. The AP1000 plant uses modern, modular-construction techniques for plant construction. The design incorporates vendor-designed skids and equipment packages, as well as large, multi-ton structural modules and special equipment modules. Modularization allows traditionally sequential construction tasks to be completed simultaneously. Factory-built modules can be installed at the site in a planned construction schedule. The modularized AP1000 plant allows many more construction activities to proceed in parallel. This reduces plant construction calendar time, thus lowering the costs of plant financing. Furthermore, performing less work onsite significantly reduces the amount of skilled field-craft labor, which costs more than shop labor. In addition to labor cost savings, doing more welding and fabrication in a factory environment raises the quality of work, allowing more scheduling flexibility and reducing the amount of specialized tools required onsite. The site layout for the AP1000 plant has been established to support modular construction and efficient operations during construction. The plant layout is compact, using less space than previous conventional plant layouts. This paper provides and overview of the AP1000 plant modules with an emphasis on structural modules. Currently the Westinghouse AP1000 plant has four units under construction in China and four units under construction in the United States. All have shown successful fabrication and installation of various AP1000 plant modules. (Author)

  9. AP1000"T"M plant modularization

    International Nuclear Information System (INIS)

    Cantarero L, C.; Demetri, K. J.; Quintero C, F. P.

    2016-09-01

    The AP1000"T"M plant is an 1100 M We pressurized water reactor (PWR) with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. Modules are used extensively in the design of the AP1000 plant nuclear island. The AP1000 plant uses modern, modular-construction techniques for plant construction. The design incorporates vendor-designed skids and equipment packages, as well as large, multi-ton structural modules and special equipment modules. Modularization allows traditionally sequential construction tasks to be completed simultaneously. Factory-built modules can be installed at the site in a planned construction schedule. The modularized AP1000 plant allows many more construction activities to proceed in parallel. This reduces plant construction calendar time, thus lowering the costs of plant financing. Furthermore, performing less work onsite significantly reduces the amount of skilled field-craft labor, which costs more than shop labor. In addition to labor cost savings, doing more welding and fabrication in a factory environment raises the quality of work, allowing more scheduling flexibility and reducing the amount of specialized tools required onsite. The site layout for the AP1000 plant has been established to support modular construction and efficient operations during construction. The plant layout is compact, using less space than previous conventional plant layouts. This paper provides and overview of the AP1000 plant modules with an emphasis on structural modules. Currently the Westinghouse AP1000 plant has four units under construction in China and four units under construction in the United States. All have shown successful fabrication and installation of various AP1000 plant modules. (Author)

  10. Discussion of QA grading for AP1000 NP plant

    International Nuclear Information System (INIS)

    Luo Shuiyun; Zhang Qingchuan

    2012-01-01

    The grading method of quality assurance for the following AP1000 project is presented based on the Westinghouse classification principle, referring to the classification method of the AP1000 self-reliance supporting project and considering the factors of classification, which can meet the requirements of domestic nuclear safety regulation and standard of the QA classification. (authors)

  11. Fuzzy uncertainty modeling applied to AP1000 nuclear power plant LOCA

    International Nuclear Information System (INIS)

    Ferreira Guimaraes, Antonio Cesar; Franklin Lapa, Celso Marcelo; Lamego Simoes Filho, Francisco Fernando; Cabral, Denise Cunha

    2011-01-01

    Research highlights: → This article presents an uncertainty modelling study using a fuzzy approach. → The AP1000 Westinghouse NPP was used and it is provided of passive safety systems. → The use of advanced passive safety systems in NPP has limited operational experience. → Failure rates and basic events probabilities used on the fault tree analysis. → Fuzzy uncertainty approach was employed to reliability of the AP1000 large LOCA. - Abstract: This article presents an uncertainty modeling study using a fuzzy approach applied to the Westinghouse advanced nuclear reactor. The AP1000 Westinghouse Nuclear Power Plant (NPP) is provided of passive safety systems, based on thermo physics phenomenon, that require no operating actions, soon after an incident has been detected. The use of advanced passive safety systems in NPP has limited operational experience. As it occurs in any reliability study, statistically non-significant events report introduces a significant uncertainty level about the failure rates and basic events probabilities used on the fault tree analysis (FTA). In order to model this uncertainty, a fuzzy approach was employed to reliability analysis of the AP1000 large break Loss of Coolant Accident (LOCA). The final results have revealed that the proposed approach may be successfully applied to modeling of uncertainties in safety studies.

  12. AP1000 design and construction integration

    International Nuclear Information System (INIS)

    Winters, James W.; Clelland, Jill A.

    2004-01-01

    Construction costs of commercial nuclear generating plants must be reduced in order to expand the future use of nuclear energy. Two of the drivers of plant construction costs are the cost of financing during the construction duration and the substantial amount of skilled craft labor hours needed on site during construction. The application of information technology (IT) has been used to understand and reduce both of these drivers by establishing parallel construction paths using modules and integrating construction sequence review into the design process. In a program sponsored by EPRI, Westinghouse has modeled the construction of AP1000 in '4D' to show its viability, to improve its logic, to improve the plant design for constructibility and overall to reduce time and risk in the construction schedule. The design of most of AP1000 was constrained to be a duplicate of AP600 except where components required expansion for the higher power level. As a result, the construction schedule for AP1000 is as mature and as robust as that for AP600. Two areas important to the construction of AP1000 did require some design work because they could not remain the same as AP1000. First, the turbine building had to be redesigned to accommodate the larger turbine and its support systems. Again, as much of the AP600 design and philosophy as possible was retained. The building required enlargement and the basemat, foundations, steel structure and structural modules required modification. As concrete, steel, and equipment were defined by the designers, they were matched to the original AP600 turbine building schedule. This forced designers to assemble files to be consistent with building assembly activities and to think about constructibility as they defined the final design. Second, the reinforcement structure within the concrete under and supporting the containment vessel required detail design. Westinghouse was fortunate to have the constructor Obayashi of Japan recommend a detailed

  13. AP1000 plant construction in China: Ansaldo Nucleare contribution

    International Nuclear Information System (INIS)

    Frogheri, Monica; Saiu, Gianfranco

    2009-01-01

    On 24th of July 2007 Westinghouse Electric Co. signed landmark contracts with China's State Nuclear Power Technology Corporation (SNPTC), to provide four AP1000 nuclear power plants in China. The AP1000 is a two-loop 1117 MWe Pressurized Water Reactor (PWR). It is based on proven technology, but with an emphasis on safety features that rely on natural driving forces, such as pressurized gas, gravity flow, natural circulation flow and convection. Ansaldo Nucleare has provided a significant support to the passive plant technology development and, starting from 2000, is cooperating with Westinghouse to development of the AP1000 Plant. In the frame of the AP1000 Chinese agreement, Ansaldo Nucleare, in Joint Venture with Mangiarotti Nuclear, has signed a contract with Westinghouse for the design and the supply of innovative components to be installed in the first AP1000 unit to be constructed at the Sanmen site. The contract includes: the design of the steel containment vessel, preparation of construction and fabrication, specifications, design and supply of SCV mechanical penetrations, air locks and equipment hatches. Moreover, Ansaldo Nucleare is in charge of the final design of the AP1000 PRHR-HX and together with Mangiarotti Nuclear will supply the component for the Sanmen Unit 1 NPP. The paper presents an overview of the design and manufacturing activities performed by Ansaldo Nucleare and its partners for the AP1000 plant in China. (authors)

  14. AP1000 - update on projects in US and China

    International Nuclear Information System (INIS)

    Godfrey, M.

    2012-01-01

    Westinghouse is the only company solely focused on commercial nuclear technology. Westinghouse business is based on four product lines regionally divided: nuclear power plants, nuclear fuel, nuclear services and nuclear automation. The AP1000 is the technology of choice for more than half of the new plants identified in the US. Westinghouse has the only certified Generation III+ technology by the US Nuclear Regulatory Commission (NRC). The first Generation III+ plants are under construction in China and the US.

  15. AP1000 - update on projects in US and China

    Energy Technology Data Exchange (ETDEWEB)

    Godfrey, M. [Westinghouse Electric Company, Cranberry Township, Pennsy lvania (United States)

    2012-07-01

    Westinghouse is the only company solely focused on commercial nuclear technology. Westinghouse business is based on four product lines regionally divided: nuclear power plants, nuclear fuel, nuclear services and nuclear automation. The AP1000 is the technology of choice for more than half of the new plants identified in the US. Westinghouse has the only certified Generation III+ technology by the US Nuclear Regulatory Commission (NRC). The first Generation III+ plants are under construction in China and the US.

  16. AP1000R licensing and deployment in the United States

    International Nuclear Information System (INIS)

    Jordan, R. P.; Russ, P. A.; Filiak, P. P.; Castiglione, L. L.

    2012-01-01

    In recent years, both domestic and foreign utilities have turned to the standardized Westinghouse AP1000 plant design in satisfying their near - and long-term - sustainable energy needs. As direct support to these actions, licensing the AP1000 design has played a significant role by providing one of the fundamental bases in clearing regulatory hurdles leading to the start of new plant construction. Within the U.S. alone, Westinghouse AP1000 licensing activities have reached unprecedented milestones with the approvals of both AP1000 Design Certification and Southern Company's combined construction permit and operating license (COL) application directly supporting the construction of two new nuclear plants in Georgia. Further COL application approvals are immediately pending for an additional two AP1000 plants in South Carolina. And, across the U.S. nuclear industry spectrum, there are 10 other COL applications under regulatory review representing some 16 new plants at 10 sites. In total, these actions represent the first wave of new plant licensing under the regulatory approval process since 1978. Fundamental to the Nuclear Regulatory Commission's AP1000 Design Certification is the formal recognition of the AP1000 passive safety design through regulatory acceptance rulemaking. Through recognition and deployment of the AP1000 Design Certification, the utility licensee / operator of this reactor design are now offered an opportunity to use a simplified 'one-step' combined license process, thereby managing substantial back-end construction schedule risk from regulatory and intervention delays. Application of this regulatory philosophy represents both acceptance and encouragement of standardized reactor designs like the AP1000. With the recent AP1000 Design Certification and utility COL acceptances, the fundamental licensing processes of this philosophy have successfully proven the attainment of significant milestones with the next stage licensing actions directed

  17. Westinghouse small modular reactor design and application

    Energy Technology Data Exchange (ETDEWEB)

    Blinn, R.; Godfrey, M. [Westinghouse Electric Company, Cranberry Township, Pennsilvania (United States)

    2012-07-01

    The AP1000 is currently under construction in both China and the US with the first one scheduled to come on line in late 2013. Nuclear power is a proven, safe, plentiful and clean source of power generation, and Westinghouse Electric Company, the pioneer and global leader in nuclear plant design and construction, is ready with the AP1000™ pressurized water reactor (PWR). The AP1000, based on the proven performance of Westinghouse-designed PWRs, is an advanced 1154 MWe nuclear power plant that uses the forces of nature and simplicity of design to enhance plant safety and operations and reduce construction costs.

  18. Finite element modeling of AP1000 nuclear island

    International Nuclear Information System (INIS)

    Tinic, S.; Orr, R.

    2003-01-01

    The AP1000 is a standard design developed by Westinghouse and its partners for an advanced nuclear power plant utilizing passive safety features. It is based on the certified design of the AP600 and has been uprated to 1000 MWe. The plant has five principal building structures; the nuclear island, the turbine building; the annex building; the diesel generator building and the radwaste building. The nuclear island consists of the containment building (the steel containment vessel and the containment internal structures), the shield building, and the auxiliary building. These structures are founded on a common basemat and are collectively known as the nuclear island. This paper describes use of the general purpose finite element program ANSYS [2] in structural analyses and qualification of the AP1000 nuclear island buildings. It describes the modeling of the shield building and the auxiliary building and the series of analyses and the flow of information from the global analyses to the detailed analyses and building qualification. (author)

  19. On integration and innovation of sino-foreign safety culture in Haiyang AP1000 Project

    International Nuclear Information System (INIS)

    Li Ruipu; Song Fengwei

    2010-01-01

    The undergoing Haiyang Nuclear Power Plant is not only introducing the top-advanced AP1000 nuclear technology, but also the mature HSE management system from U.S.A. It's very important for both sides to communicate, comprehend and acculturation of both different culture. After over 1 year discussion and practice, the experts of Westinghouse Consortium and Chinese HSE engineers have established an distinctive safety culture of AP1000 Project initially, demonstrating the followings: Exemplary actions of the expat experts and the SNPTC leaders, the high level standard HSE procedures, HSE audit, various training, HSE inspection all-around, the safety performance assessment for prospective index, JHA/JSA , emergency system, humanism rewards and punishment etc.. Haiyang SPMO has made Three-Step master plan for AP1000 project HSE Routine by analysis the site problems and the difference between Chinese and American, that is, from 2008 to 2020, when nuclear power achieve to independent, safety culture of Haiyang AP1000 will change from 'dependent' to 'independent', until the last 'interdependent'. (authors)

  20. The AP1000{sup R} China projects move forward to construction completion and equipment installation

    Energy Technology Data Exchange (ETDEWEB)

    Harrop, G. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    The AP1000 design is the only Generation III+ technology to receive design certification from the U.S. Nuclear Regulatory Commission. This evolutionary design provides the highest safety and performance standards and has several distinct advantages over other designs, including improved operations and reduced construction schedule risks through the use of modern, modular, engineering principles that allow construction and fabrication tasks traditionally performed in sequence to be undertaken in parallel. Since the first granting of Design Certification in 2005 by the NRC, the AP1000 design has been modified to meet emergent NRC requirements such as those requiring the design to withstand the impact of an aircraft crash. Both domestic and foreign utilities have turned to the Westinghouse AP1000 plant design to meet their near - and long-term sustainable energy needs. The first ever deployment of this advanced U.S. nuclear power technology began in China in 2007 with the award of a contract to build four AP1000 units, constructed in pairs at the coastal sites of Sanmen (Zhejiang Province) and Haiyang (Shandong Province). Currently, all four units are at an advanced stage of construction. The commercial operation date for Sanmen Unit 1 is November 2013 followed by Haiyang Unit 1 being operational in May 2014. Construction and equipment manufacture is at an advanced stage. Sanmen Unit 1 equipment that has been delivered includes the reactor vessel, the reactor vessel closure head, the passive residual heat removal heat exchanger, the integrated head package, the polar crane, and the refueling machine. The steam generators are also completed. The RV was installed within the containment vessel building in September 2011. The installation of this major equipment will allow the setting of the containment vessel top head. Haiyang Unit 1 is also achieving significant progress. Significant benefits continue to be realized as a result of lessons learned and experience gained

  1. AP1000 will meet the challenges of near-term deployment

    International Nuclear Information System (INIS)

    Matzie, Regis A.

    2008-01-01

    The world demand for energy is growing rapidly, particularly in developing countries that are trying to raise the standard of living for billions of people, many of whom do not have access to electricity or clean water. Climate change and the concern for increased emissions of green house gases have brought into question the future primary reliance of fossil fuels. With the projected worldwide increase in energy demand, concern for the environmental impact of carbon emissions, and the recent price volatility of fossil fuels, nuclear energy is undergoing a rapid resurgence. This 'nuclear renaissance' is broad based, reaching across Asia, North America, Europe, as well as selected countries in Africa and South America. Many countries have publicly expressed their intentions to pursue the construction of new nuclear energy plants. Some countries that have previously turned away from commercial nuclear energy are reconsidering the advisability of this decision. This renaissance is facilitated by the availability of more advanced reactor designs than are operating today, with improved safety, economy, and operations. One such design, the Westinghouse AP1000 advanced passive plant, has been a long time in the making! The development of this passive technology started over two decades ago from an embryonic belief that a new approach to design was needed to spawn a nuclear renaissance. The principal challenges were seen as ensuring reactor safety by requiring less reliance on operator actions and overcoming the high plant capital cost of nuclear energy. The AP1000 design is based on the use of innovative passive technology and modular construction, which require significantly less equipment and commodities that facilitate a more rapid construction schedule. Because Westinghouse had the vision and the perseverance to continue the development of this passive technology, the AP1000 design is ready to meet today's challenge of near-term deployment

  2. AP1000: Meeting economic goals in a competitive world. Annex 7

    International Nuclear Information System (INIS)

    Davis, G.; Cummins, E.; Winters, J.

    2002-01-01

    In the U.S., conditions are becoming more favorable for considering the nuclear option again for new baseload generation. While oil and natural gas prices have risen, the cost of operating the existing fleet of nuclear plants has decreased. Furthermore, an advanced 1000 MWe nuclear plant that will be even more cost-competitive with fossil fuels and natural gas will be available by 2005. Westinghouse, in an effort to further improve on the AP600's cost competitiveness, has developed the AP1000, a two-loop, 1000 MWe, advanced pressurized water reactor (PWR) with passive safety features and extensive plant simplifications to enhance the construction, operation, and maintenance. Like the AP600, the AP1000 uses proven technology that builds on over 30 years of operating PWR experience. Westinghouse has completed design studies that demonstrate that it is feasible to increase the power output of the AP600 to at least 1000 MWe, maintaining its current design configuration and licensing basis. To maximize the cost savings, the AP1000 has been designed within the space constraints of the AP600, while retaining the credibility of proven components and substantial safety margins. The affect on the plant's overnight cost of the increased major components that is required to uprate the AP600 to 1000 MWe is small. This overall cost addition is on the order of 11 percent, while the overall power increase is almost 80 percent. This paper describes the changes made to uprate the AP600 and gives an overview of the plant design. (author)

  3. Westinghouse AP600 advanced nuclear plant design

    International Nuclear Information System (INIS)

    Gangloff, W.

    1999-01-01

    As part of the cooperative US Department of Energy (DOE) Advanced Light Water Reactor (ALWR) Program and the Electric Power Research Institute (EPRI), the Westinghouse AP600 team has developed a simplified, safe, and economic 600-megawatt plant to enter into a new era of nuclear power generation. Designed to satisfy the standards set by DOE and defined in the ALWR Utility Requirements Document (URD), the Westinghouse AP600 is an elegant combination of innovative safety systems that rely on dependable natural forces and proven technologies. The Westinghouse AP600 design simplifies plant systems and significant operation, inspections, maintenance, and quality assurance requirements by greatly reducing the amount of valves, pumps, piping, HVAC ducting, and other complex components. The AP600 safety systems are predominantly passive, depending on the reliable natural forces of gravity, circulation, convection, evaporation, and condensation, instead of AC power supplies and motor-driven components. The AP600 provides a high degree of public safety and licensing certainty. It draws upon 40 years of experience in light water reactor components and technology, so no demonstration plant is required. During the AP600 design program, a comprehensive test program was carried out to verify plant components, passive safety systems components, and containment behavior. When the test program was completed at the end of 1994, the AP600 became the most thoroughly tested advanced reactor design ever reviewed by the US Nuclear Regulatory Commission (NRC). The test results confirmed the exceptional behavior of the passive systems and have been instrumental in facilitating code validations. Westinghouse received Final Design Approval from the NRC in September 1998. (author)

  4. Characteristics and design improvement of AP1000 automatic depressurization system

    International Nuclear Information System (INIS)

    Jin Fei

    2012-01-01

    Automatic depressurization system, as a specialty of AP1000 Design, enhances capability of mitigating design basis accidents for plant. Advancement of the system is discussed by comparing with traditional PWR design and analyzing system functions, such as depressurizing and venting. System design improvement during China Project performance is also described. At the end, suggestions for the system in China Project are listed. (author)

  5. Operating experience review for the AP1000 plant

    International Nuclear Information System (INIS)

    Chaney, T. E.; Lipner, M. H.

    2006-01-01

    Westinghouse is performing an update to the Operating Experience Review (OER) Report for the AP1000 project to account for operating experience since December 1996. Significant Operating Experience Reports, Significant Event Reports, Significant Event Notifications, Operations and Maintenance Reminders, Topical Reports, Event Analysis Reports and Licensee Event Reports were researched for pertinent input to the update. As a part of the OER, Westinghouse has also conducted operator interviews and observations during simulated plant operations and after operating events. The main purpose of the OER is to identify Human Factors Engineering (HFE) related safety issues from existing operating plant experience and to ensure that these issues are addressed in the new design. The issues and lessons learned regarding operating experience provide a basis for improving the plant design. (authors)

  6. AP1000{sup R} licensing and deployment in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, R. P.; Russ, P. A.; Filiak, P. P.; Castiglione, L. L. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    In recent years, both domestic and foreign utilities have turned to the standardized Westinghouse AP1000 plant design in satisfying their near - and long-term - sustainable energy needs. As direct support to these actions, licensing the AP1000 design has played a significant role by providing one of the fundamental bases in clearing regulatory hurdles leading to the start of new plant construction. Within the U.S. alone, Westinghouse AP1000 licensing activities have reached unprecedented milestones with the approvals of both AP1000 Design Certification and Southern Company's combined construction permit and operating license (COL) application directly supporting the construction of two new nuclear plants in Georgia. Further COL application approvals are immediately pending for an additional two AP1000 plants in South Carolina. And, across the U.S. nuclear industry spectrum, there are 10 other COL applications under regulatory review representing some 16 new plants at 10 sites. In total, these actions represent the first wave of new plant licensing under the regulatory approval process since 1978. Fundamental to the Nuclear Regulatory Commission's AP1000 Design Certification is the formal recognition of the AP1000 passive safety design through regulatory acceptance rulemaking. Through recognition and deployment of the AP1000 Design Certification, the utility licensee / operator of this reactor design are now offered an opportunity to use a simplified 'one-step' combined license process, thereby managing substantial back-end construction schedule risk from regulatory and intervention delays. Application of this regulatory philosophy represents both acceptance and encouragement of standardized reactor designs like the AP1000. With the recent AP1000 Design Certification and utility COL acceptances, the fundamental licensing processes of this philosophy have successfully proven the attainment of significant milestones with the next stage licensing

  7. Westinghouse Advances in Passive Plant Safety

    International Nuclear Information System (INIS)

    Bruschi, H. J.; Manager, General; Gerstenhaber, E.

    1993-01-01

    On June 26, 1992, Westinghouse submitted the Ap600 Standard Safety Analysis Report and comprehensive PIRA results to the U. S. NRC for review as part of the Ap600 design certification program. This major milestone was met on time on a schedule set more than 3 years before submittal and is the result of the cooperative efforts of the U. S. Department of Energy (DOE), the Electric Power Requirements Program, and the Westinghouse Ap600 design team. These efforts were initiated in 1985 to develop a 600 MW advanced light water reactor plant design based on specific technical requirements established to provide the safety, simplicity, reliability, and economics necessary for the next generation of nuclear power plants. The Ap600 design achieves the ALRR safety requirements through ample design margins, simplified safety systems based on natural driving forces, and on a human-engineered man-machine interface system. Extensive Probabilistic Risk evolution, have recently shown that even if none of the active defense-in-depth safety systems are available, the passive systems alone meet safety goals. Furthermore, many tests in an extensive test program have begun or have been completed. Early tests show that passive safety perform well and meet design expectations

  8. Significant advantages of the safety-first concept in construction, operation, and maintenance of the Westinghosue AP1000 reactor

    International Nuclear Information System (INIS)

    Cummins, E.; Benitz, K.

    2004-01-01

    In June 2003, the U.S. Nuclear Regulatory Commission (USNRC) published a draft opinion about safety of the AP1000 Westinghouse pressurized water reactor with 'passive safety' features. The report constitutes an important milestone in the development of the next generation of safe and cost-efficient nuclear power plants. A new AP1000 can be absolutely competitive with fossil fired power plants and may be able to revive the construction of new nuclear power plants worldwide. The reason for designing the AP1000 were safety considerations. The use of passive safety systems at the same time entails a considerable reduction in the costs of design, maintenance, and operation of an AP1000 plant. Independent experts confirmed that an AP1000 can be erected within three years or even less. The estimated electricity generating costs of an AP1000 plant in the United States amount to US Cent 3.2 to 3.6 per kilowatthour. (orig.)

  9. Difference of reactor core nuclear instrument between AP1000 and CPR1000

    International Nuclear Information System (INIS)

    Zhang Shidong; Zhou Can; Deng Tian

    2014-01-01

    As a typical generation Ⅲ reactor technique, the AP1000 applies many advanced design concepts, simplifies the design, reduces equipment quantities, and thus enhances systematic reliability. The comparison of reactor core measurement instrument differences between AP1000 and CPR1000 from several aspects was involved in the paper. Through analysis and comparison of these differences, passive design concepts and characteristics of AP1000 are familiarized, and conveniences for staffs engaged in CPR1000 to learn and grasp AP1000 technique are provided. It is useful in reactor start up, operation and maintenance. (authors)

  10. Introduction to sump screen downstream effect analysis of AP1000 nuclear power plant

    International Nuclear Information System (INIS)

    Zhang Qinghua; Liu Yu; Chai Guohan

    2010-01-01

    The design of AP1000 takes into account the potential impact of debris clogging on sump screen. In this article, the technical background of sump screen issue and the design characteristics of AP1000 to address the sump screen blockage issue are introduced. The article focuses on the 'downstream effect' analysis method, acceptance criteria and analysis result of AP1000 sump screen. Although the design of AP1000 is different with traditional PWR, the author expects to bring some reference to advance the downstream effect analysis in China through the introduction. (authors)

  11. Dynamic Response of AP1000 Nuclear Island Due to Safe Shutdown Earthquake Loading

    Directory of Open Access Journals (Sweden)

    Gan Buntara S.

    2017-01-01

    Full Text Available AP1000 is a standard nuclear power plant developed by Westinghouse and its partners by using an advanced passive safety feature. Among the five principle building structures, namely the nuclear island, turbine building, annex building, diesel generator building and radwaste building, the safety of the nuclear island building is the most concerned. This paper investigates the dynamic response of the nuclear island building of the AP1000 plant subjected to safe shutdown earthquake loadings. A finite element model for the building, which is assumed to be built in a hard-rock base, is developed and its dynamic response is computed with the aid of the commercial finite element package ANSYS. The dynamic characteristics, including the natural frequencies, the vibration modes, and the time histories for displacements, velocities, and accelerations of the building are obtained for two typical safe shutdown earthquakes, El Centro and Kobe earthquakes. The dynamic behavior of the building due to the earthquakes and its safety is examined and highlighted.

  12. Application of computerized procedure system for AP1000

    International Nuclear Information System (INIS)

    Wen Fang

    2012-01-01

    With the tendency of digitalisation in instrumentation and control system of nuclear power plants, AP1000, as well as other advanced PWRs, is provided with the technical basis for the application of CPS (computerized procedure system). This paper makes a brief introduction on CPS construction and function. CPS, as an advanced procedure system, does not only have the function of electronic indication for operation procedures, but also have the ability to monitor plant data, process the data and then present the status of the procedure steps to the reactor operator. In addition, advantages of CPS compared with paper-based procedures and comparison with digital operating system of one M310 plus advanced nuclear power plant are described. Moreover, based on current situation, this paper offers several suggestions on CPS localization for Sanmen AP1000 nuclear power project. Besides, the last part of this paper discusses problems we might meet during the process of CPS localization. (author)

  13. AP1000 Containment Design and Safety Assessment

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  14. OLP embedment design method research for AP1000 nuclear plant

    International Nuclear Information System (INIS)

    Li Cheng; Li Shaoping; Liu Jianwei

    2013-01-01

    Background: One of the most advanced nuclear power technology, the first AP1000 reactor is under construction in China. Modularization is one of the main characteristics for AP1000 nuclear plant building. Module wall with steel face plate is used instead of reinforced concrete structure wall. A number of OLP embedments need to be installed into the module wall to connect other structures such as pipes, equipment, operation platforms and any other component attached to the module wall. Therefore, the design of embedment is very important in AP1000 structural design. Purpose: A finite element analysis method and tool for embedment design is needed for convenience. Methods: This paper applies the self-developed GTStrudl command template and VBA macro program for embedment capacity calculation and evaluation based on Microsoft Excel to the embedment design. Results: A Microsoft Excel template for embedment design is developed. Conclusions: The analysis method and template brings reasonable results and may provide some help and use for reference for the engineering practice. (authors)

  15. Structural modules in AP1000 plant design

    International Nuclear Information System (INIS)

    Prasad, N.; Tunon-Sanjur, L.

    2007-01-01

    Structural modules are extensively used in AP1000 plant design. The shop manufacturing of modules components improves the quality and reliability of plant structures. The application of modules has a positive impact on construction schedules, and results in substantial savings in the construction cost. This paper describes various types of structural modules used for AP1000 plant structures. CA structural wall modules are steel plate modules with concrete placed, on or within the module, after module installation. The layout and design of the largest CA wall modules, CA01 and CA20, is described in detail. General discussion of structural floor modules, such as the composite and finned floors, is also included. Steel form CB modules (liners) consist of plate reinforced with angle stiffeners and tee sections. The angles and the tee sections are on the concrete side of the plate. Design of CB20 has been included as an example of CB type modules. Design codes and structural concepts related to module designs are discussed. (authors)

  16. AP1000 shield building: a constructability challenge

    International Nuclear Information System (INIS)

    Di Giuseppe, Giovanni; Bonanno, Domenico

    2010-01-01

    The AP1000 Shield Building, an enhanced structure which surrounds the containment vessel, consists of standard Reinforced Concrete (RC) and composite Steel and Concrete (SC) construction. In the SC module the surface steel plates, (with attached shear studs and angles) filled with concrete, act as the steel reinforcement in concrete. This is a relatively new design technology that required the appropriate use of structural codes, supplemented with information from applicable tests on similar composite steel and concrete construction. Being a newer design concept, existing codes do not provide explicit guidance on SC construction so a review of literature and test data on composite structures similar to AP1000 shield building was done in order to confirm the technical basis for the design. The SC walls, air inlet structure and roof of the Shield Building will be constructed using modular construction practices and then transported to site and lifted into place. These modules, working also as permanent form-work, will be filled with high strength Self- Consolidating Concrete. (SCC) This paper provides a focused and integrated presentation of the enhanced shield building design methodology, testing, constructability and inspection. (authors)

  17. Study of impact of the AP1000{sup Registered-Sign} reactor vessel upper internals design on fuel performance

    Energy Technology Data Exchange (ETDEWEB)

    Xu Yiban; Conner, Michael; Yuan Kun; Dzodzo, Milorad B.; Karoutas, Zeses; Beltz, Steven A.; Ray, Sumit; Bissett, Teresa A. [Westinghouse Electric Company, Cranberry Township, PA 16066 (United States); Chieng, Ching-Chang, E-mail: cchieng@ess.nthu.edu.tw [National Tsing Hua University, Hsinchu 30043, Taiwan (China); Kao, Min-Tsung; Wu, Chung-Yun [National Tsing Hua University, Hsinchu 30043, Taiwan (China)

    2012-11-15

    One aspect of the AP1000{sup Registered-Sign} reactor design is the reduction in the number of major components and simplification in manufacturing. One design change relative to current Westinghouse reactors of similar size is the reduction in the number of reactor vessel outlet nozzles/hot legs leaving the upper plenum from three to two. With regard to fuel performance, this design difference creates a different flow field in the AP1000 reactor vessel upper plenum (the region above the core). The flow exiting core and entering the upper plenum must turn 90 Degree-Sign , flow laterally through the upper plenum around support structures, and exit through one of the two outlet nozzles. While the flow in the top of the core is mostly axial, there is some lateral flow component as the core flow reacts to the flow field and pressure distribution in the upper plenum. The pressure distribution in the upper plenum varies laterally depending upon various factors including the proximity to the outlet nozzles. To determine how the lateral flow in the top of the AP1000 core compares to current Westinghouse reactors, a computational fluid dynamics (CFD) model of the flow in the upper portion of the AP1000 reactor vessel including the top region of the core, the upper plenum, the reactor vessel outlet nozzles, and a portion of the hot legs was created. Due to geometric symmetry, the computational domain was reduced to a quarter (from the top view) that includes Vulgar-Fraction-One-Quarter of the top of the core, Vulgar-Fraction-One-Quarter of the upper plenum, and Vulgar-Fraction-One-Half of an outlet nozzle. Results from this model include predicted velocity fields and pressure distributions throughout the model domain. The flow patterns inside and around guide tubes clearly demonstrate the influence of lateral flow due to the presence of the outlet nozzles. From these results, comparisons of AP1000 flow versus current Westinghouse plants were performed. Field performance

  18. Study on time management of FCD concrete pouring AP1000 NPP and its application

    International Nuclear Information System (INIS)

    Wu Jie; Yang Ming; Cong Jiuyuan

    2010-01-01

    Haiyang nuclear power plant adopts the advanced third-generation nuclear power technology, AP1000, the design concept of passive system and the methods of modularization and 'open-top' construction greatly shortens the construction period. At the same time, higher requirements are put forward to the management of construction, quality and progress control. This paper will apply the statistical process control theory and method to the time management of FCD (First Concrete Day) to Unit 1 of Haiyang AP1000 nuclear power plant, and it brings up the mathematical model of time management based on verification through modeling, data analysis, model optimization and the actual construction work. The theory and method studied in this paper can not be only applied to the FCD concrete pouring for Unit 1 of Haiyang AP1000 nuclear power plant, but also have great referential and guiding significance to the continuous concreting of the mass concrete of the follow-up similar construction. (authors)

  19. ANALISIS KONDISI TERAS REAKTOR DAYA MAJU AP1000 PADA KECELAKAAN SMALL BREAK LOCA

    Directory of Open Access Journals (Sweden)

    Andi Sofrany Ekariansyah

    2015-06-01

    , mixture level, temperatur kelongsong, small break LOCA, RELAP5.   ABSTRACT ANALYSIS ON THE CORE CONDITION OF AP1000 ADVANCED POWER REACTOR DURING SMALL BREAK LOCA. Accident due to the loss of coolant from the reactor boundary is an anticipated design basis event in the design of power reactor adopting the Generation II up to IV technology. Small break LOCA leads to more significant impact on safety compared to the large break LOCA as shown in the Three-Mile Island (TMI. The focus of this paper is the small break LOCA analysis on the Generation III+ advanced power reactor of AP1000 by simulating three different initiating events, which are inadvertent opening of Automatic Depressurization System (ADS, double-ended break on one of Direct Vessel Injection (DVI pipe, and 10 inch diameter split break on one of cold leg pipe. Methodology used is by simulating the events on the AP1000 model developed using RELAP5/SCDAP/Mod3.4. The impact analyzed is the core condition during the small break LOCA consisting of the mixture level occurrence and the fuel cladding temperature transient. The results show that the mixture level for all small break LOCA events are above the active core height, which indicates no core uncovery event. The development of the mixture level affect the fuel cladding temperature transient, which shows a decreasingly trend after the break, and the effectifeness of core cooling. Those results are identical with the results of other code of NOTRUMP. The resulted core cooling is also due to the function of coolant injection from passive safety feature, which is unique in the AP1000 design. In overall, the result of analysis shows that the AP1000 model developed by the RELAP5 can be used for analysis of design basis accident considered in the AP1000 advanced power reactor. Keywords: analysis, mixture level, fuel cladding temperature, small break LOCA, RELAP5.

  20. Control parameter optimization for AP1000 reactor using Particle Swarm Optimization

    International Nuclear Information System (INIS)

    Wang, Pengfei; Wan, Jiashuang; Luo, Run; Zhao, Fuyu; Wei, Xinyu

    2016-01-01

    Highlights: • The PSO algorithm is applied for control parameter optimization of AP1000 reactor. • Key parameters of the MSHIM control system are optimized. • Optimization results are evaluated though simulations and quantitative analysis. - Abstract: The advanced mechanical shim (MSHIM) core control strategy is implemented in the AP1000 reactor for core reactivity and axial power distribution control simultaneously. The MSHIM core control system can provide superior reactor control capabilities via automatic rod control only. This enables the AP1000 to perform power change operations automatically without the soluble boron concentration adjustments. In this paper, the Particle Swarm Optimization (PSO) algorithm has been applied for the parameter optimization of the MSHIM control system to acquire better reactor control performance for AP1000. System requirements such as power control performance, control bank movement and AO control constraints are reflected in the objective function. Dynamic simulations are performed based on an AP1000 reactor simulation platform in each iteration of the optimization process to calculate the fitness values of particles in the swarm. The simulation platform is developed in Matlab/Simulink environment with implementation of a nodal core model and the MSHIM control strategy. Based on the simulation platform, the typical 10% step load decrease transient from 100% to 90% full power is simulated and the objective function used for control parameter tuning is directly incorporated in the simulation results. With successful implementation of the PSO algorithm in the control parameter optimization of AP1000 reactor, four key parameters of the MSHIM control system are optimized. It has been demonstrated by the calculation results that the optimized MSHIM control system parameters can improve the reactor power control capability and reduce the control rod movement without compromising AO control. Therefore, the PSO based optimization

  1. AP1000 construction and operating costs

    Energy Technology Data Exchange (ETDEWEB)

    Winters, J.W.; Corletti, M.M. [Westinghouse Electric Corp., Pittsburgh, PA (United States); Thompson, M

    2001-07-01

    Market analysis of the U.S. electricity generating market indicates that the generating cost of competitive new generating capacity must be less than $0.03/kw-hr. When such factors as an attractive return on investment and payback period are considered for a new nuclear electric generating facility, this results in the requirement to have an overnight capital cost of approximately $1000/kw. Industry executives indicate that any new nuclear plant must be able to compete in the de-regulated generation wholesale marketplace and provide a return to the shareholders. Against this standard, the costs of advanced nuclear power plants currently available are still too high. In the United States, the Utility Requirements Document for advanced light water reactor plants included a cost goal that was based on the cost of coal generated electricity at the time the document was written. Since that time, the cost of new generating capacity and the overall operating cost of generating electricity has gone down. This is a result of low natural gas prices, more efficient plants in general and the current record breaking reductions in outage times and operating costs for nuclear plants. The plant designs resulting from the United States advanced light water reactor plant programs received Design Certification from the United States Nuclear Regulatory Commission. Some are being deployed, with variations, in countries other than the United States. But they can not compete today with other sources of central station generation in the United States. (author)

  2. AP1000 construction and operating costs

    International Nuclear Information System (INIS)

    Winters, J.W.; Corletti, M.M.; Thompson, M.

    2001-01-01

    Market analysis of the U.S. electricity generating market indicates that the generating cost of competitive new generating capacity must be less than $0.03/kw-hr. When such factors as an attractive return on investment and payback period are considered for a new nuclear electric generating facility, this results in the requirement to have an overnight capital cost of approximately $1000/kw. Industry executives indicate that any new nuclear plant must be able to compete in the de-regulated generation wholesale marketplace and provide a return to the shareholders. Against this standard, the costs of advanced nuclear power plants currently available are still too high. In the United States, the Utility Requirements Document for advanced light water reactor plants included a cost goal that was based on the cost of coal generated electricity at the time the document was written. Since that time, the cost of new generating capacity and the overall operating cost of generating electricity has gone down. This is a result of low natural gas prices, more efficient plants in general and the current record breaking reductions in outage times and operating costs for nuclear plants. The plant designs resulting from the United States advanced light water reactor plant programs received Design Certification from the United States Nuclear Regulatory Commission. Some are being deployed, with variations, in countries other than the United States. But they can not compete today with other sources of central station generation in the United States. (author)

  3. AP1000 passive core cooling system pre-operational tests procedure definition and simulation by means of Relap5 Mod. 3.3 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Lioce, D., E-mail: donato.lioce@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Asztalos, M., E-mail: asztalmj@westinghouse.com [Westinghouse Electric Company, Cranberry Twp, PA 16066 (United States); Alemberti, A., E-mail: alessandro.alemberti@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Barucca, L. [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Frogheri, M., E-mail: monicalinda.frogheri@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy); Saiu, G., E-mail: gianfranco.saiu@aen.ansaldo.it [Ansaldo Nucleare S.p.A., Corso F. M. Perrone 25, 16161, Genova (Italy)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Two AP1000 Core Make-up Tanks pre-operational tests procedures have been defined. Black-Right-Pointing-Pointer The two tests have been simulated by means of the Relap5 computer code. Black-Right-Pointing-Pointer Results show the tests can be successfully performed with the selected procedures. - Abstract: The AP1000{sup Registered-Sign} plant is an advanced Pressurized Water Reactor designed and developed by Westinghouse Electric Company which relies on passive safety systems for core cooling, containment isolation and containment cooling, and maintenance of main control room emergency habitability. The AP1000 design obtained the Design Certification by NRC in January 2006, as Appendix D of 10 CFR Part 52, and it is being built in two locations in China. The AP1000 plant will be the first commercial nuclear power plant to rely on completely passive safety systems for core cooling and its licensing process requires the proper operation of these systems to be demonstrated through some pre-operational tests to be conducted on the real plant. The overall objective of the test program is to demonstrate that the plant has been constructed as designed, that the systems perform consistently with the plant design, and that activities culminating in operation at full licensed power including initial fuel load, initial criticality, and power increase to full load are performed in a controlled and safe manner. Within this framework, Westinghouse Electric Company and its partner Ansaldo Nucleare S.p.A. have strictly collaborated, being Ansaldo Nucleare S.p.A. in charge of the simulation of some pre-operational tests and supporting Westinghouse in the definition of tests procedures. This paper summarizes the work performed at Ansaldo Nucleare S.p.A. in collaboration with Westinghouse Electric Company for the Core Makeup Tank (CMT) tests, i.e. the CMTs hot recirculation test and the CMTs draindown test. The test procedure for the two

  4. AP1000 passive core cooling system pre-operational tests procedure definition and simulation by means of Relap5 Mod. 3.3 computer code

    International Nuclear Information System (INIS)

    Lioce, D.; Asztalos, M.; Alemberti, A.; Barucca, L.; Frogheri, M.; Saiu, G.

    2012-01-01

    Highlights: ► Two AP1000 Core Make-up Tanks pre-operational tests procedures have been defined. ► The two tests have been simulated by means of the Relap5 computer code. ► Results show the tests can be successfully performed with the selected procedures. - Abstract: The AP1000 ® plant is an advanced Pressurized Water Reactor designed and developed by Westinghouse Electric Company which relies on passive safety systems for core cooling, containment isolation and containment cooling, and maintenance of main control room emergency habitability. The AP1000 design obtained the Design Certification by NRC in January 2006, as Appendix D of 10 CFR Part 52, and it is being built in two locations in China. The AP1000 plant will be the first commercial nuclear power plant to rely on completely passive safety systems for core cooling and its licensing process requires the proper operation of these systems to be demonstrated through some pre-operational tests to be conducted on the real plant. The overall objective of the test program is to demonstrate that the plant has been constructed as designed, that the systems perform consistently with the plant design, and that activities culminating in operation at full licensed power including initial fuel load, initial criticality, and power increase to full load are performed in a controlled and safe manner. Within this framework, Westinghouse Electric Company and its partner Ansaldo Nucleare S.p.A. have strictly collaborated, being Ansaldo Nucleare S.p.A. in charge of the simulation of some pre-operational tests and supporting Westinghouse in the definition of tests procedures. This paper summarizes the work performed at Ansaldo Nucleare S.p.A. in collaboration with Westinghouse Electric Company for the Core Makeup Tank (CMT) tests, i.e. the CMTs hot recirculation test and the CMTs draindown test. The test procedure for the two selected tests has been defined and, in order to perform the pre-operational tests simulations, a

  5. AP1000 - the new standard for nuclear power

    International Nuclear Information System (INIS)

    Lipman, Daniel S.

    2006-01-01

    Full text of publication follows: The AP1000 is the only Generation III+ reactor to receive Final Design Approval (FDA) from the Nuclear Regulatory Commission, and is expected to receive its Design Certification by the end of the year. Building on the proven features of current generation nuclear plants, the AP1000 combines experience with innovation into a design that surpasses current standards of safety and reliability. Use of passive safety features results in a simpler and more compact design that enhances safety, simplifies O and M requirements, and reduces capital and operating costs. At 1117 Mwe, the AP1000 is well suited for almost any grid system and will be fully competitive with combined-cycle gas and comparable fossil fuel plants. The AP1000 is ready to help launch a renaissance in new nuclear plant construction throughout the world. Maturity of Design: In excess of 1300 man-years and $400 million in development funding have been expended on the AP1000. It has undergone extensive, part scale testing at the system, sub-system and component level, in addition to a series of part scale integrated tests. The AP1000 is the most analyzed of the next generation reactors. Simplicity of Design/Economics: The AP1000 uses simplified and innovative passive safety systems to an unprecedented extent. Simplified passive safety systems provide reliable operation, reduced capital costs, and enhanced plant safety with large plant operating margins. The AP1000 features improved reliability through simplicity rather than addition of redundant safety trains. This simpler design is easier and less costly to operate and maintain than larger, more complex plants, while less equipment and smaller buildings translate into lower capital costs and shorter construction durations. After construction, economic benefit will be found in reduced operating and maintenance costs, largely due to reduced operating and maintenance staffing requirements. Construction aspects

  6. European utility requirements (EUR) volume 3 assessment for AP1000

    International Nuclear Information System (INIS)

    Saiu, G.; Demetri, K.J.

    2005-01-01

    The EUR (European Utility Requirements) Volume 3 is intended to report the Plant Description, the Compliance Assessment to EUR Volumes 1 and 2, and finally, the Specific Requirements for each specific Nuclear Power Plant Design considered by the EUR. Five subsets of EUR Volume 3, based on EUR Revision B, are already published; all of which are next generation plant designs being developed for Europe beyond 2000. They include : 1) EP1000 - Passive Pressurized Light Water Reactor (3-Loop, 1000 MWe) 2) EPR - Evolutionary Pressurized Light Water Reactor (1500 MWe) 3) BWR90/90+ - Evolutionary Boiling Water Reactor (1400 MWe) 4) ABWR - Evolutionary Boiling Water Reactor (1400 MWe) 5) SWR 1000 - Boiling Water Reactor With Passive Features (1000 MWe) In addition, the following subsets are currently being developed: 1) AP1000 - Passive Pressurized Light Water Reactor (2-Loop, 1117 MWe) 2) VVER AES 92 - Pressurized Water Reactor With Passive Features (1000 MWe) The purpose of this paper is to provide an overview of the program, which started in January 2004 with the EUR group to prepare an EUR Volume 3 Subset for the AP1000 nuclear plant design. The AP1000 EUR compliance assessment, to be performed against EUR Revision C requirements, is an important step for the evaluation of the AP1000 design for application in Europe. The AP1000 compliance assessment is making full use of AP1000 licensing documentation, EPP Phase 2 design activities and EP1000 EUR detailed compliance assessment. As of today, nearly all of the EUR Chapters have been discussed within the EUR Coordination Group. Based on the results of the compliance assessment, it can be stated that the AP1000 design shows a good level of compliance with the EUR Revision C requirements. Nevertheless, the compliance assessment has highlighted areas for where the AP1000 plant deviates from the EUR. The EPP design group has selected the most significant ones for performing detailed studies to quantify the degree of compliance

  7. RELAP5/SCDAPSIM model development for AP1000 and verification for large break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Trivedi, A.K. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Allison, C. [Innovative Systems Software, Idaho Falls, ID 83406 (United States); Khanna, A., E-mail: akhanna@iitk.ac.in [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Munshi, P. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India)

    2016-08-15

    Highlights: • RELAP5/SCDAPSIM model of AP1000 has been developed. • Analysis involves a LBLOCA (double ended guillotine break) study in cold leg. • Results are compared with those of WCOBRA–TRAC and TRACE. • Concluded that PCT does not violate the safety criteria of 1477 K. - Abstract: The AP1000 is a Westinghouse 2-loop pressurized water reactor (PWR) with all emergency core cooling systems based on natural circulation. Its core design is very similar to a 3-loop PWR with 157 fuel assemblies. Westinghouse has reported their results of the safety analysis in its design control document (DCD) for a large break loss of coolant accident (LOCA) using WCOBRA/TRAC and for a small break LOCA using NOTRUMP. The current study involves the development of a representative RELAP5/SCDASIM model for AP1000 based on publically available data and its verification for a double ended cold leg (DECL) break in one of the cold legs in the loop containing core makeup tanks (CMT). The calculated RELAP5/SCDAPSIM results have been compared to publically available WCOBRA–TRAC and TRACE results of DECL break in AP1000. The objective of this study is to benchmark thermal hydraulic model for later severe accident analyses using the 2D SCDAP fuel rod component in place of the RELAP5 heat structures which currently represent the fuel rods. Results from this comparison provides sufficient confidence in the model which will be used for further studies such as a station blackout. The primary circuit pumps, pressurizer and steam generators (including the necessary secondary side) are modeled using RELAP5 components following all the necessary recommendations for nodalization. The core has been divided into 6 radial rings and 10 axial nodes. For the RELAP5 thermal hydraulic calculation, the six groups of fuel assemblies have been modeled as pipe components with equivalent flow areas. The fuel including the gap and cladding is modeled as a 1d heat structure. The final input deck achieved

  8. Application of fault tree methodology to modeling of the AP1000 plant digital reactor protection system

    International Nuclear Information System (INIS)

    Teolis, D.S.; Zarewczynski, S.A.; Detar, H.L.

    2012-01-01

    The reactor trip system (RTS) and engineered safety features actuation system (ESFAS) in nuclear power plants utilizes instrumentation and control (IC) to provide automatic protection against unsafe and improper reactor operation during steady-state and transient power operations. During normal operating conditions, various plant parameters are continuously monitored to assure that the plant is operating in a safe state. In response to deviations of these parameters from pre-determined set points, the protection system will initiate actions required to maintain the reactor in a safe state. These actions may include shutting down the reactor by opening the reactor trip breakers and actuation of safety equipment based on the situation. The RTS and ESFAS are represented in probabilistic risk assessments (PRAs) to reflect the impact of their contribution to core damage frequency (CDF). The reactor protection systems (RPS) in existing nuclear power plants are generally analog based and there is general consensus within the PRA community on fault tree modeling of these systems. In new plants, such as AP1000 plant, the RPS is based on digital technology. Digital systems are more complex combinations of hardware components and software. This combination of complex hardware and software can result in the presence of faults and failure modes unique to a digital RPS. The United States Nuclear Regulatory Commission (NRC) is currently performing research on the development of probabilistic models for digital systems for inclusion in PRAs; however, no consensus methodology exists at this time. Westinghouse is currently updating the AP1000 plant PRA to support initial operation of plants currently under construction in the United States. The digital RPS is modeled using fault tree methodology similar to that used for analog based systems. This paper presents high level descriptions of a typical analog based RPS and of the AP1000 plant digital RPS. Application of current fault

  9. Confirmatory analysis of the AP1000 passive residual heat removal heat exchanger with 3-D computational fluid dynamic analysis

    International Nuclear Information System (INIS)

    Schwall, James R.; Karim, Naeem U.; Thakkar, Jivan G.; Taylor, Creed; Schulz, Terry; Wright, Richard F.

    2006-01-01

    The AP1000 is an 1100 MWe advanced nuclear power plant that uses passive safety features to enhance plant safety and to provide significant and measurable improvements in plant simplification, reliability, investment protection and plant costs. The AP1000 received final design approval from the US-NRC in 2004. The AP1000 design is based on the AP600 design that received final design approval in 1999. Wherever possible, the AP1000 plant configuration and layout was kept the same as AP600 to take advantage of the maturity of the design and to minimize new design efforts. As a result, the two-loop configuration was maintained for AP1000, and the containment vessel diameter was kept the same. It was determined that this significant power up-rate was well within the capability of the passive safety features, and that the safety margins for AP1000 were greater than those of operating PWRs. A key feature of the passive core cooling system is the passive residual heat removal heat exchanger (PRHR HX) that provides decay heat removal for postulated LOCA and non-LOCA events. The PRHR HX is a C-tube heat exchanger located in the in-containment refueling water storage tank (IRWST) above the core promoting natural circulation heat removal between the reactor cooling system and the tank. Component testing was performed for the AP600 PRHR HX to determine the heat transfer characteristics and to develop correlations to be used for the AP1000 safety analysis codes. The data from these tests were confirmed by subsequent integral tests at three separate facilities including the ROSA facility in Japan. Owing to the importance of this component, an independent analysis has been performed using the ATHOS-based computational fluid dynamics computer code PRHRCFD. Two separate models of the PRHR HX and IRWST have been developed representing the ROSA test geometry and the AP1000 plant geometry. Confirmation of the ROSA test results were used to validate PRHRCFD, and the AP1000 plant model

  10. Standardized safety management of AP1000 nuclear power plant

    International Nuclear Information System (INIS)

    Li Xingwen; Cao Zhiqiang; Cong Jiuyuan

    2011-01-01

    In 2002, China published and implemented the Law of the People's Republic of China on Work Safety and promulgated a series of guidelines and policies, which strengthened the safety management supervision. Standardization of safety, as another important step on safety supervision, comes after safety assesment and safety production licensing system, is also a permanent solution. Standardization of safety is a strategic, long term and fundamental work, which is also the basic access to achieving scientific safety management and increasing the inherent safety of an enterprise. Haiyang AP1000 nuclear power plant, adopting the modularized, 'open-top' and parallel construction means, overturned the traditional construction theory of installation work comes after the civil work and greatly shorten the construction period. At the same time, the notable increase of oversize module transportation and lifting and parallel construction raises higher demands for safety management. This article combines the characteristics and difficulties of safety management for Haiyang AP1000 nuclear power plant, puts forward ideas and methods for standardized safety management, and could also serve as reference to the safety management for other AP1000 projects. (authors)

  11. Addressing the fundamental issues in reliability evaluation of passive safety of AP1000 for a comparison with active safety of PWR

    International Nuclear Information System (INIS)

    Hashim Muhammad; Yoshikawa, Hidekazu; Yang Ming

    2013-01-01

    Passive safety systems adopted in advanced Pressurized Water Reactor (PWR), such as AP1000 and EPR, should attain higher reliability than the existing active safety systems of the conventional PWR. The objective of this study is to discuss the fundamental issues relating to the reliability evaluation of AP1000 passive safety systems for a comparison with the active safety systems of conventional PWR, based on several aspects. First, comparisons between conventional PWR and AP1000 are made from the both aspects of safety design and cost reduction. The main differences between these PWR plants exist in the configurations of safety systems: AP1000 employs the passive safety system while reducing the number of active systems. Second, the safety of AP1000 is discussed from the aspect of severe accident prevention in the event of large break loss of coolant accidents (LOCA). Third, detailed fundamental issues on reliability evaluation of AP1000 passive safety systems are discussed qualitatively by using single loop models of safety systems of both PWRs plants. Lastly, methodology to conduct quantitative estimation of dynamic reliability for AP1000 passive safety systems in LOCA condition is discussed, in order to evaluate the reliability of AP1000 in future by a success-path-based reliability analysis method (i.e., GO-FLOW). (author)

  12. Westinghouse Advanced Doped Pellet - Characteristics and irradiation behavior

    International Nuclear Information System (INIS)

    Backman, K.; Hallstadius, L.; Roennberg, G.

    2009-01-01

    Full text: There are a number of trends in the nuclear power industry, which put additional requirements on the operational flexibility and reliability of nuclear fuel, for example power uprates and longer cycles in order to increase production, higher burnup levels in order to reduce the backend cost of the fuel cycle, and lower goals for activity release from power plant operation. These additional requirements can be addressed by increasing the fuel density, improving the FG retention, improving the PCI resistance and improving the post-failure performance. In order to achieve that, Westinghouse has developed ADOPT (Advanced Doped Pellet Technology) UO 2 fuel containing additions of chromium and aluminium oxides. The additives facilitate pellet densification during sintering, enlarge the pellet grain size, and increase the creep rate. The final manufactured doped pellets reach about 0.5 % higher density within a shorter sintering time and a five times larger grain size compared with standard UO 2 fuel pellets. Fuel rods with ADOPT pellets have been irradiated in several light water reactors (LWRs) since 1999, including two full SVEA Optima2 reloads in 2005. ADOPT pellets has been investigated in pool-side and hot cell Post Irradiation Examinations (PIEs), as well as in a ramp test and a fuel washout test in the Studsvik R2 test reactor. The investigations have identified three areas of improved operational behaviour: Reduced Fission Gas Release (FGR), improved Pellet Cladding Interaction (PCI) performance thanks to increased pellet plasticity and higher resistance against post-failure degradation. The better FGR behaviour of ADOPT has been verified with a pool side FGR gamma measurement performed at 55 MWd/kgU, as well as transient tests in the Studsvik R2 reactor. Creep measurements performed on fresh pellets show that ADOPT has a higher creep rate which is beneficial for the PCI performance. ADOPT has also been part of a high power Halden test (IFA-677). The

  13. Analysis of an accident type sbloca in reactor contention AP1000 with 8.0 Gothic code; Analisis de un accidente tipo Sbloca en la contencion del reactor AP1000 con el codigo Gothic 8.0

    Energy Technology Data Exchange (ETDEWEB)

    Goni, Z.; Jimenez Varas, G.; Fernandez, K.; Queral, C.; Montero, J.

    2016-08-01

    The analysis is based on the simulation of a Small Break Loss-of-Coolant-Accident in the AP1000 nuclear reactor using a Gothic 8.0 tri dimensional model created in the Science and Technology Group of Nuclear Fision Advanced Systems of the UPM. The SBLOCA has been simulated with TRACE 5.0 code. The main purpose of this work is the study of the thermo-hydraulic behaviour of the AP1000 containment during a SBLOCA. The transients simulated reveal close results to the realistic behaviour in case of an accident with similar characteristics. The pressure and temperature evolution enables the identification of the accident phases from the RCS point of view. Compared to the licensing calculations included in the AP1000 Safety Analysis, it has been proved that the average pressure and temperature evolution is similar, yet lower than the licensing calculations. However, the temperature and inventory distribution are significantly heterogeneous. (Author)

  14. Engineering human factors into the Westinghouse advanced control room

    International Nuclear Information System (INIS)

    Easter, J.R.

    1987-01-01

    By coupling the work of the Riso Laboratory in Denmark on human behaviour with new digital computation and display technology, Westinghouse has developed a totally new control room design. This design features a separate, co-ordinated work station to support the systems management role in decision making, as well as robust alarm and display systems. This coupling of the functional and physical data presentation is now being implemented in test facilities. (author)

  15. Quantitative dynamic reliability evaluation of AP1000 passive safety systems by using FMEA and GO-FLOW methodology

    International Nuclear Information System (INIS)

    Hashim Muhammad; Yoshikawa, Hidekazu; Matsuoka, Takeshi; Yang Ming

    2014-01-01

    The passive safety systems utilized in advanced pressurized water reactor (PWR) design such as AP1000 should be more reliable than that of active safety systems of conventional PWR by less possible opportunities of hardware failures and human errors (less human intervention). The objectives of present study are to evaluate the dynamic reliability of AP1000 plant in order to check the effectiveness of passive safety systems by comparing the reliability-related issues with that of active safety systems in the event of the big accidents. How should the dynamic reliability of passive safety systems properly evaluated? And then what will be the comparison of reliability results of AP1000 passive safety systems with the active safety systems of conventional PWR. For this purpose, a single loop model of AP1000 passive core cooling system (PXS) and passive containment cooling system (PCCS) are assumed separately for quantitative reliability evaluation. The transient behaviors of these passive safety systems are taken under the large break loss-of-coolant accident in the cold leg. The analysis is made by utilizing the qualitative method failure mode and effect analysis in order to identify the potential failure mode and success-oriented reliability analysis tool called GO-FLOW for quantitative reliability evaluation. The GO-FLOW analysis has been conducted separately for PXS and PCCS systems under the same accident. The analysis results show that reliability of AP1000 passive safety systems (PXS and PCCS) is increased due to redundancies and diversity of passive safety subsystems and components, and four stages automatic depressurization system is the key subsystem for successful actuation of PXS and PCCS system. The reliability results of PCCS system of AP1000 are more reliable than that of the containment spray system of conventional PWR. And also GO-FLOW method can be utilized for reliability evaluation of passive safety systems. (author)

  16. AP1000 core design with 50% MOX loading

    International Nuclear Information System (INIS)

    Fetterman, Robert J.

    2009-01-01

    The European uility requirements (EUR) document states that the next generation European passive plant (EPP) reactor core design shall be optimized for UO 2 fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO 2 core design and a mixed MOX/UO 2 core design, discussing relevant results related to reactivity management, power margin and fuel rod performance

  17. AP1000 core design with 50% MOX loading

    Energy Technology Data Exchange (ETDEWEB)

    Fetterman, Robert J. [Westinghouse Electric Company, LLC, Pittsburgh, PA (United States)

    2008-07-01

    The European Utility Requirements (EUR) document states that the next generation European Passive Plant (EPP) reactor core design shall be optimized for UO{sub 2} fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO{sub 2} core and a mixed MOX / UO{sub 2} core design, discussing relevant results related to reactivity management, power margin and fuel rod performance. (authors)

  18. AP1000 core design with 50% MOX loading

    International Nuclear Information System (INIS)

    Fetterman, Robert J.

    2008-01-01

    The European Utility Requirements (EUR) document states that the next generation European Passive Plant (EPP) reactor core design shall be optimized for UO 2 fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO 2 core and a mixed MOX / UO 2 core design, discussing relevant results related to reactivity management, power margin and fuel rod performance. (authors)

  19. AP1000 core design with 50% MOX loading

    Energy Technology Data Exchange (ETDEWEB)

    Fetterman, Robert J. [Westinghouse Electric Company, LLC, Pittsburgh, PA (United States)], E-mail: fetterrj@westinghouse.com

    2009-04-15

    The European uility requirements (EUR) document states that the next generation European passive plant (EPP) reactor core design shall be optimized for UO{sub 2} fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO{sub 2} core design and a mixed MOX/UO{sub 2} core design, discussing relevant results related to reactivity management, power margin and fuel rod performance.

  20. Seismic analysis for the supporting member of the Westinghouse AP1000 steam generator

    International Nuclear Information System (INIS)

    Xu Yu; Huang Mei; Tian Li; Hou Zhousen

    2012-01-01

    In this paper, the seismic performance analysis for the Supporting member of is carried out under the combined loads, including dead weight, earthquake loads, by using response spectrum analysis method in ANSYS. The stress qualification is also carried out based on ASME-Ⅲ-NF code. The results show that the stress of the Supporting member meets the seismic requirements for equipment, and the deformation of structure is within the allowable limits. (authors)

  1. Characterization of liquid entrainment in the AP1000 automatic depressurization system from APEX tests

    International Nuclear Information System (INIS)

    Richard F Wright; Terry L Schulz; Jose N Reyes; John Groome

    2005-01-01

    Full text of publication follows: The AP1000 is a 1000 MWe advanced nuclear power plant that uses passive safety features to enhance plant safety and to provide significant and measurable improvements in plant simplification, reliability, investment protection and plant costs. The AP1000 relies heavily on the 600 MWe AP600 which received design certification in 1999. A critical part of the AP600 design certification process involved the testing of the passive safety systems. A one-fourth height, one-fourth pressure test facility, APEX-600, was constructed at the Oregon State University to study design basis events, and to provide a body of data to be used to validate the computer models used to analyze the AP600. This facility was extensively modified to reflect the design changes for AP1000 including higher power in the electrically heated rods representing the reactor core, and changes in the size of the pressurizer, core makeup tanks and automatic depressurization system. The APEX-1000 test facility was used to perform design basis accident simulations and separate effects tests to support the AP1000 design certification process. In the event of a LOCA, the AP1000 passive core cooling system provides sources of core makeup water along with an automatic depressurization system (ADS) consisting of several stages of valves which reduce the reactor coolant system pressure in a controlled manner. The final stage of this system, ADS-4, consists of four large valves that open off the hot legs, reducing the pressure to allow gravity injection from the in-containment refueling water storage tank (IRWST) and eventually the containment sump. The 67% increase in power from AP600 to AP1000 results in proportionally larger steam velocities exiting the core. Higher steam velocities could increases the potential for significant liquid entrainment out the ADS-4 lines, affecting the liquid inventory in the reactor. Tests were performed in APEX-1000 to characterize the two

  2. Westinghouse employs advanced robotics in a state-of-the-art LWR line

    International Nuclear Information System (INIS)

    Anon.

    1985-01-01

    To increase productivity while maintaining quality, Westinghouse's new Manufacturing Automation Process for oxide fuel features Integrated Dry Route conversion technology, a fully-integrated management information system, advanced robotics and enhanced materials handling practices. The new line is expected to begin operating in 1985. (author)

  3. Westinghouse employs advanced robotics in a state-of-the-art LWR line

    Energy Technology Data Exchange (ETDEWEB)

    1985-03-01

    To increase productivity while maintaining quality, Westinghouse's new Manufacturing Automation Process for oxide fuel features Integrated Dry Route conversion technology, a fully-integrated management information system, advanced robotics and enhanced materials handling practices. The new line is expected to begin operating in 1985.

  4. The study on neutron and photon distribution of AP1000 reactor by MCNP code

    International Nuclear Information System (INIS)

    Chen Defeng; Shen Mingqi

    2014-01-01

    The core and reactor structural of AP1000 was modeled by the MCNP calculation program which is based on the Monte Carlo method in this paper, the neutron and photon distribution of AP1000 reactor core was calculated by the conditions of reactor critical. The results show that the AP1000 reactor neutron and photon distribution is in accordance with the critical design of PWR. (authors)

  5. Comparison and analysis for item classifications between AP1000 and traditional PWR

    International Nuclear Information System (INIS)

    Luo Shuiyun; Liu Xiaoyan

    2012-01-01

    The comparison and analysis for the safety classification, seismic category, code classification and QA classification between AP1000 and traditional PWR were presented. The safety could be guaranteed and the construction and manufacture costs could be cut down since all sorts of AP1000 classifications. It is suggested that the QA classification and the QA requirements correspond to the national conditions should be drafted in the process of AP1000 domestication. (authors)

  6. Influence of performance improvement of AP1000 nuclear island main equipment forging on manufacturing

    International Nuclear Information System (INIS)

    Liu Zhiying

    2013-01-01

    In order to comply with the 60-year design life of an AP1000 nuclear power station, higher strength and ductility requirements have been made on AP1000 nuclear island SG forgings than on CPR1000 nuclear island main equipment. In addition, bigger size of AP1000 nuclear island SG forgings increases the difficulty of manufacturing them. Insufficient recognition of these changes may cause unstable quality of forgings and possible quality problems in follow-up welding procedure. On the basis of comparison and analysis of AP1000 nuclear island SG forgings and CPR1000 nuclear island forgings, this thesis suggests clear directions for the actions we need to take. (author)

  7. Large LOCA accident analysis for AP1000 under earthquake

    International Nuclear Information System (INIS)

    Yu, Yu; Lv, Xuefeng; Niu, Fenglei

    2015-01-01

    Highlights: • Seismic failure event probability is induced by uncertainties in PGA and in Am. • Uncertainty in PGA is shared by all the components at the same place. • Relativity induced by sharing PGA value can be analyzed explicitly by MC method. • Multi components failures and accident sequences will occur under high PGA value. - Abstract: Seismic probabilistic safety assessment (PSA) is developed to give the insight of nuclear power plant risk under earthquake and the main contributors to the risk. However, component failure probability including the initial event frequency is the function of peak ground acceleration (PGA), and all the components especially the different kinds of components at same place will share the common ground shaking, which is one of the important factors to influence the result. In this paper, we propose an analysis method based on Monte Carlo (MC) simulation in which the effect of all components sharing the same PGA level can be expressed by explicit pattern. The Large LOCA accident in AP1000 is analyzed as an example, based on the seismic hazard curve used in this paper, the core damage frequency is almost equal to the initial event frequency, moreover the frequency of each accident sequence is close to and even equal to the initial event frequency, while the main contributors are seismic events since multi components and systems failures will happen simultaneously when a high value of PGA is sampled. The component failure probability is determined by uncertainties in PGA and in component seismic capacity, and the former is the crucial element to influence the result

  8. Westinghouse technologies and integration with Toshiba

    International Nuclear Information System (INIS)

    Noda, Tetsuya; Tanazawa, Takeshi; Yoshida, Hiroyuki

    2007-01-01

    With Westinghouse Electric Company (WEC) now a member of the Toshiba Group, Toshiba is capable of supplying both boiling water reactor (BWR) and pressurized water reactor (PWR) systems. WEC is well experienced worldwide in the nuclear business and by integrating the technologies of both Toshiba and WEC. Toshiba will be able to provide a greater range of services in the global market. We will build a cooperative structure not only for the maintenance service and fuel businesses but also for the development of innovative reactors while aiming for global expansion with the AP 1000 PWR, the most advanced PWR in the nuclear power plant business. We will continue making efforts so as to be able to provide all types of products and services as one-stop solutions regardless of the type of reactor. (author)

  9. The Westinghouse AP600 an advanced nuclear option for small or medium electricity grids

    International Nuclear Information System (INIS)

    Bruschi, H. J.; Novak, V.

    1996-01-01

    During the early days of commercial nuclear power, many countries looking to add nuclear power to their energy mix required large plants to meet the energy needs of rapidly growing populations and large industrial complexes. The majority of plants worldwide are in the range of 100 megawatts and beyond. During the 1970s, it became apparent that a smaller nuclear plants would appeal to utilities looking to add additional power capacity to existing grids, or to utilities in smaller countries which were seeking efficient, new nuclear generation capacity for the first time. For instance, the Westinghouse-designed 600 megawatt Krsko plant in Slovenia began operation in 1980, providing electricity to inhabitants of relatively small, yet industrial populations of Slovenia and Croatia. This plant design incorporated the best, proven technology available at that time, based on 20 years of Westinghouse PWR pioneering experience. Beginning in the early 1980s, Westinghouse began to build further upon that experience - in part through the advanced light water reactor programs established by the Electric Power Research institute (EPRI) and the U.S. Department of Energy (DOE) - to design a simplified, advanced nuclear reactor in the 600 megawatt range. Originally, Westinghouse's development of its AP600 (advanced, passive 600-megawatt) plants was geared towards the needs of U.S. utilities which specified smaller, simplified nuclear options for the decades ahead. It soon became evident that the small and medium sized electricity grids of international markets could benefit from this new reactor. From the earliest days of Westinghouse's AP600 development, the corporation invited members of the international nuclear community to take part in the design, development and testing of the AP600 - with the goal of designing a reactor that would meet the diverse needs of an international industry composed of countries with similar, yet different, concerns. (author)

  10. Neutronic and thermal-hydraulic calculations for the AP-1000 NPP with the MCNP6 and SERPENT codes

    International Nuclear Information System (INIS)

    Stefani, Giovanni Laranjo; Maiorino, Jose R.; Santos, Thiago A.

    2015-01-01

    The AP-1000 is an evolutionary PWR reactor designed as an evolution of the AP-600 project. The reactor is already pre-licensed by NRC, and is considered to have achieved high standards of safety, possible short construction time and good economic competitiveness. The core is a 17x17 typical assembly using Zirlo as cladding, 3 different enrichment regions, and is controlled by boron, control banks, and burnable poison. The expected fuel final burnup is 62 MWD/ton U and a cycle of 18 months. In this paper we present results for neutronic and thermal-hydraulic calculations for the AP-1000. We use the MCNP6 and SERPENT codes to calculate the first cycle of operation. The calculated parameters are K eff at BOL and EOL and its variation with burnup and neutron flux, and reactivity coefficients. The production of transuranic elements such as Pu-239 and Pu-241, and burning fuel are calculated over time. In the work a complete reactor was burned for 450 days with no control elements, boron or burnable poison were considered, these results were compared with data provided by the Westinghouse. The results are compared with those reported in the literature. A simple thermal hydraulic analysis allows verification of thermal limits such as fuel and cladding temperatures, and MDNB. (author)

  11. Neutronic and thermal-hydraulic calculations for the AP-1000 NPP with the MCNP6 and SERPENT codes

    Energy Technology Data Exchange (ETDEWEB)

    Stefani, Giovanni Laranjo; Maiorino, Jose R.; Santos, Thiago A., E-mail: giovanni.laranjo@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br, E-mail: thiago.santos@ufabc.edu.br [Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais; Rossi, Pedro R., E-mail: pedro.russorossi@gmail.com [FERMIUM - Tecnologia Nuclear, Sao Paulo, SP (Brazil)

    2015-07-01

    The AP-1000 is an evolutionary PWR reactor designed as an evolution of the AP-600 project. The reactor is already pre-licensed by NRC, and is considered to have achieved high standards of safety, possible short construction time and good economic competitiveness. The core is a 17x17 typical assembly using Zirlo as cladding, 3 different enrichment regions, and is controlled by boron, control banks, and burnable poison. The expected fuel final burnup is 62 MWD/ton U and a cycle of 18 months. In this paper we present results for neutronic and thermal-hydraulic calculations for the AP-1000. We use the MCNP6 and SERPENT codes to calculate the first cycle of operation. The calculated parameters are K{sub eff} at BOL and EOL and its variation with burnup and neutron flux, and reactivity coefficients. The production of transuranic elements such as Pu-239 and Pu-241, and burning fuel are calculated over time. In the work a complete reactor was burned for 450 days with no control elements, boron or burnable poison were considered, these results were compared with data provided by the Westinghouse. The results are compared with those reported in the literature. A simple thermal hydraulic analysis allows verification of thermal limits such as fuel and cladding temperatures, and MDNB. (author)

  12. Research on quality assurance classification methodology for domestic AP1000 nuclear power projects

    International Nuclear Information System (INIS)

    Bai Jinhua; Jiang Huijie; Li Jingyan

    2012-01-01

    To meet the quality assurance classification requirements of domestic nuclear safety codes and standards, this paper analyzes the quality assurance classification methodology of domestic AP1000 nuclear power projects at present, and proposes the quality assurance classification methodology for subsequent AP1000 nuclear power projects. (authors)

  13. MDEP AP1000WG Design-Specific Common Position CP-AP1000WG-02. Common position addressing Fukushima Daiichi NPP accident-related issues

    International Nuclear Information System (INIS)

    2016-09-01

    A severe accident involving several units took place in Japan at Fukushima Daiichi nuclear power plant (NPP) in March 2011. The immediate cause of the accident was an earthquake followed by a tsunami coupled with inadequate provisions against the consequences of such events in the design. Opportunities to improve protection against a realistic design basis tsunami had not been taken. As a consequence of the tsunami, safety equipment and the related safety functions were lost at the plant, leading to core damage in three units and subsequently to large radioactive release. Several studies have already been performed to better understand the accident progression and detailed technical studies are still in progress in Japan and elsewhere. In the meantime, on-going studies on the behaviour of nuclear power plants in very severe situations, similar to Fukushima Daiichi, seek to identify potential vulnerabilities in plant design and operation; to suggest reasonably practicable upgrades; or to recommend enhanced regulatory requirements and guidance to address such situations. Likewise, agencies around the world that are responsible for regulating the design, construction and operation of AP1000 R plants are engaged in similar activities. The MDEP AP1000 R Working Group (AP1000 WG) members consist of members from Canada, China, the United Kingdom and the United States. Since the regulatory review of their AP1000 R applications have not been completed by all of these Countries yet, this paper identifies common preliminary approaches to address potential safety improvements for AP1000 R plants as related to lessons learned from the Fukushima Daiichi accident or Fukushima Daiichi-related issues. In seeking common position, regulators will provide input to this paper to reflect their safety conclusions regarding the AP1000 R design and how the design could be enhanced to address Fukushima Daiichi issues. The common preliminary approaches are organized into five sections

  14. Enhancing AP1000 reactor accident management capabilities for long term accidents

    International Nuclear Information System (INIS)

    Jiang Pingting; Liu Mengying; Duan Chengjie; Liao Yehong

    2015-01-01

    Passive safety actions are considered as main measures under severe accident in AP1000 power plant. However, risk is still existed. According to PSA, several probable scenarios for AP1000 nuclear power plant are analyzed in this paper with MAAP the severe accident analysis code. According to the analysis results, several deficiencies of AP1000 severe accident management are found. The long term cooling and containment depressurization capability for AP1000 power plant appear to be most important factors under such accidents. Then, several temporary strategies for AP1000 power plant are suggested, including PCCWST temporary water supply strategy after 72h, temporary injection strategy for IRWST, hydrogen relief action in fuel building, which would improve the safety of AP1000 power plant. At last, assessments of effectiveness for these strategies are performed, and the results are compared with analysis without these strategies. The comparisons showed that correct actions of these strategies would effectively prevent the accident process of AP1000 power plant. (author)

  15. Dynamic Analysis of AP1000 Shield Building Considering Fluid and Structure Interaction Effects

    Directory of Open Access Journals (Sweden)

    Qiang Xu

    2016-02-01

    Full Text Available The shield building of AP1000 was designed to protect the steel containment vessel of the nuclear reactor. Therefore, the safety and integrity must be ensured during the plant life in any conditions such as an earthquake. The aim of this paper is to study the effect of water in the water tank on the response of the AP1000 shield building when subjected to three-dimensional seismic ground acceleration. The smoothed particle hydrodynamics method (SPH and finite element method (FEM coupling method is used to numerically simulate the fluid and structure interaction (FSI between water in the water tank and the AP1000 shield building. Then the grid convergence of FEM and SPH for the AP1000 shield building is analyzed. Next the modal analysis of the AP1000 shield building with various water levels (WLs in the water tank is taken. Meanwhile, the pressure due to sloshing and oscillation of the water in the gravity drain water tank is studied. The influences of the height of water in the water tank on the time history of acceleration of the AP1000 shield building are discussed, as well as the distributions of amplification, acceleration, displacement, and stresses of the AP1000 shield building. Research on the relationship between the WLs in the water tank and the response spectrums of the structure are also taken. The results show that the high WL in the water tank can limit the vibration of the AP1000 shield building and can more efficiently dissipate the kinetic energy of the AP1000 shield building by fluid-structure interaction.

  16. Study on the operation mode for indigenization and standardization of AP1000 technology

    International Nuclear Information System (INIS)

    Gao Zhihu; Cheng Huiping

    2014-01-01

    This paper describes the importance and necessity of developing standardized AP1000 technology, and analyzes the problems faced and measures to be taken. The operation mode, known as the Committee of AP1000 Standardized Design and Innovation Management, was first put forward in China. And the paper also discusses how to arouse the enthusiasm of the owners, designers, manufactures, as well as construction, installation, regulation and other parties, how to exploit the advantage of whole industry to promote the localization and standardization of AP1000 technology. (authors)

  17. Comparison of CPR1000 and AP1000 rod position indication systems

    International Nuclear Information System (INIS)

    Lei Qing

    2009-01-01

    This paper introduces the structure, the function, the digital detection principle of reactor control rod position and monitoring systems in CPR1000 and AP1000, comparing with the characteristics of the system design. The results show that the operation mode and function of AP1000 Rod position indication system are similar to that of CPR1000, but AP1000 rod position system provides higher reliability, and reduces the numbers of containment electrical penetrations and is with better characteristics than that of CPR1000, since it incorporated the redundancy design and data communication. (authors)

  18. Characteristics and application study of AP1000 NPPs equipment reliability classification method

    International Nuclear Information System (INIS)

    Guan Gao

    2013-01-01

    AP1000 nuclear power plant applies an integrated approach to establish equipment reliability classification, which includes probabilistic risk assessment technique, maintenance rule administrative, power production reliability classification and functional equipment group bounding method, and eventually classify equipment reliability into 4 levels. This classification process and result are very different from classical RCM and streamlined RCM. It studied the characteristic of AP1000 equipment reliability classification approach, considered that equipment reliability classification should effectively support maintenance strategy development and work process control, recommended to use a combined RCM method to establish the future equipment reliability program of AP1000 nuclear power plants. (authors)

  19. HSE management for AP1000 nuclear plant construction in EPC mode

    International Nuclear Information System (INIS)

    He Xiaogang; Wei Zhong

    2010-01-01

    As a new nuclear type, AP1000 will become the development direction of Chinese nuclear project. EPC General Contract mode is favored by nuclear owners both at home and abroad. Therefore, there is necessity for studying HSE management system and method suitable for AP1000 nuclear plant construction (ANPC) based on combination of AP1000 construction characters in EPC mode. This can not only ensure safety for ANPC but also positively promote national nuclear power development. For this reason, based on site HSE management of the first AP1000 nuclear plant under construction, HSE management system and method for ANPC in EPC mode was proposed after analysis of the character of EPC mode and ANPC character. It is hoped that it will be helpful for safe construction for ANPC. (authors)

  20. The modularization construction of piping system installation in AP1000 plant

    International Nuclear Information System (INIS)

    Lu Song; Wang Yuan; Wei Junming

    2012-01-01

    Modularization construction is the main technique used in AP1000 plants, the piping Modularization installation will impact directly to the module construction as the important part of the Modularization construction. After the piping system has took the modularization design in AP1000 plants, some installation works of piping system has moved from the site to fabrication shop. With improving the construction quality and minimizing the time frame of project, the critical paths can be optimized. This paper has analyzed the risk and challenge that met during the modularization construction period of piping systems though introducing the characteristic of modularization construction for AP1000 piping systems, and get construction experiences from the First AP1000 plants in the world, then it will be the firmly basics for the wide application of modularization construction in the future. (authors)

  1. Analysis of risk management during AP1000 equipment technology transfer and localization

    International Nuclear Information System (INIS)

    Gao Yongjun; Guan Rui

    2009-01-01

    This article analyzes the risk factors existing in AP1000 equipment technology transfer and localization process by describing the invitation for bid, tender evaluation and contract negotiation process of the third-generation nuclear power plant technology introduction project of China and discusses the classification, evaluation and analysis methods of risks, and puts forward some referential suggestions for the successful introduction of equipment technology for AP1000 nuclear project. (authors)

  2. Comparison Study of Water Demineralization System for the OPR 1000 and AP 1000 Nuclear Power Plant

    International Nuclear Information System (INIS)

    Dedy Priambodo; Siti Alimah; Erlan Dewita

    2009-01-01

    OPR 1000 adopts demineralization method based on ion exchanger resin and AP 1000 adopt the method that based on Reverse Osmosis (RO)-Electrodeionization (EDI). The Ion exchange process is a reversible chemical reaction of a solution and an insoluble solid. Ion exchanger use resin as polluter ions capture and will be regenerated after its saturated. RO is method using pressure to force a solution through a membrane, retaining the solute on one side and allowing the pure solvent to pass to the other side. Whereas, EDI is a combination of ion exchange and electrodialysis. The ions is taken by ion exchange resin, and then it is discharged utilizing electric potential difference. Due to water splitting phenomena in EDI, make resin will never be saturated, so the RO-EDI process is water demineralization system that use little chemical, more simple installation, capable to maintain demineralization water product quality and environmental friendly. Thereby, The RO-EDI water demineralization system is more advance then ion exchange technology. (author)

  3. Development of an advanced 16x165 Westinghouse type PWR fuel assembly for Slovenia

    International Nuclear Information System (INIS)

    Boone, M. L.; King, S. J.; Pulver, E. F.; Jeon, K.-L.; Esteves, R.; Kurincic, B.

    2004-01-01

    Industrias Nucleares do Brasil (INB), KEPCO Nuclear Fuel Company, Ltd. (KNFC), and Westinghouse Electric Company (Westinghouse) have jointly designed an advanced 16x16 Westinghouse type PWR fuel assembly. This advanced 16x16 Westinghouse type PWR fuel assembly, which will be implemented in both Kori Unit 2 (in Korea) and Angra Unit 1 (in Brazil) in January and March 2005, respectively, is an integral part of the utilities fuel management strategy. This same fuel design has also been developed for future use in Krsko Unit 1 (in Slovenia). In this paper we will describe the front-end nuclear fuel management activities utilized by the joint development team and describe how these activities played an integral part in defining the direction of the advanced 16x16 Westinghouse type PWR fuel assembly design. Additionally, this paper will describe how this design demonstrates improved margins under high duty plant operating conditions. The major reason for initiating this joint development program was to update the current 16x16 fuel assembly, which is also called 16STD. The current 16STD fuel assembly contains a non-optimized fuel rod diameter for the fuel rod pitch (i.e. 9.5 mm OD fuel rods at a 0.485 inch pitch), non-neutronic efficient components (i.e. Inconel Mid grids), no Intermediate Flow Mixer (IFM) grids, and other mechanical features. The advanced 16x16 fuel assembly is being designed for peak rod average burnups of up to 75 MWd/kgU and will use an optimized fuel rod diameter (i.e. 9.14 mm OD ZIRLO TM fuel rods), neutronic efficient components (i.e. ZIRLO TM Mid grids), ZIRLO TM Intermediate Flow Mixer (IFM) grids to improve Departure from Nucleate Boiling (DNB) margin, and many other mechanical features that improve design margins. Nuclear design activities in the areas of fuel cycle cost and fuel management were performed in parallel to the fuel assembly design efforts. As the change in reactivity due to the change in the fuel rod diameter influences directly

  4. Study on modular construction management in AP1000 nuclear plant project

    International Nuclear Information System (INIS)

    Fang Xiaopeng; Shen Wenrong; Sun Kebin; Wei Zhong

    2010-01-01

    The construction of AP1000 Nuclear Power Plant (NPP) has commenced in China. The AP1000 NPP features a passive design concept and modular construction technology. Based on the management of the construction of current AP1000 NNP, this paper describes the effects on Nuclear Island (NI) construction project management resulting from modular construction technology, as well as new construction techniques and methods. This paper puts forward new requirements for construction schedule management of the nuclear island construction at different levels. The AP1000 NI construction logic features the parallel construction of civil and structural erection as the main approach, with the integrated schedule of module fabrication, assembly and installation as support. The structural modules of AP1000 project are prefabricated in shop, delivered to site as sub-modules and assembled to integrated structural module. The assembled module is transported to the construction site, hoisted and finally set in NI. This paper illustrates how to ensure the construction quality of structural modules by analyzing the interface process and key links in the quality control program, and introduces how to ensure the safety of heavy structural components during various construction phases by evaluating and analyzing the construction safety process. This paper also makes an analysis of the safe environment for the assembly and installation of Containment Vessel, the management of product protection and personnel safety inside the Containment Building during 'Open Top' construction, raises to implement effective protection for the numerous pre-set mechanical modules and equipments, as well as personnel safety protection programs and measures. The modular construction feature of AP1000 NPP design requires technique improvement and management innovation during the NI construction. This paper makes a study and research on the control management of schedule, quality and safety of AP1000 NPP NI

  5. Simulation and Analysis of Small Break LOCA for AP1000 Using RELAP5-MV and Its Comparison with NOTRUMP Code

    Directory of Open Access Journals (Sweden)

    Eltayeb Yousif

    2017-01-01

    Full Text Available Many reactor safety simulation codes for nuclear power plants (NPPs have been developed. However, it is very important to evaluate these codes by testing different accident scenarios in actual plant conditions. In reactor analysis, small break loss of coolant accident (SBLOCA is an important safety issue. RELAP5-MV Visualized Modularization software is recognized as one of the best estimate transient simulation programs of light water reactors (LWR. RELAP5-MV has new options for improved modeling methods and interactive graphics display. Though the same models incorporated in RELAP5/MOD 4.0 are in RELAP5-MV, the significant difference of the latter is the interface for preparing the input deck. In this paper, RELAP5-MV is applied for the transient analysis of the primary system variation of thermal hydraulics parameters in primary loop under SBLOCA in AP1000 NPP. The upper limit of SBLOCA (10 inches is simulated in the cold leg of the reactor and the calculations performed up to a transient time of 450,000.0 s. The results obtained from RELAP5-MV are in good agreement with those of NOTRUMP code obtained by Westinghouse when compared under the same conditions. It can be easily inferred that RELAP5-MV, in a similar manner to RELAP5/MOD4.0, is suitable for simulating a SBLOCA scenario.

  6. Discrete rod burnup analysis capability in the Westinghouse advanced nodal code

    International Nuclear Information System (INIS)

    Buechel, R.J.; Fetterman, R.J.; Petrunyak, M.A.

    1992-01-01

    Core design analysis in the last several years has evolved toward the adoption of nodal-based methods to replace traditional fine-mesh models as the standard neutronic tool for first core and reload design applications throughout the nuclear industry. The accuracy, speed, and reduction in computation requirements associated with the nodal methods have made three-dimensional modeling the preferred approach to obtain the most realistic core model. These methods incorporate detailed rod power reconstruction as well. Certain design applications such as confirmation of fuel rod design limits and fuel reconstitution considerations, for example, require knowledge of the rodwise burnup distribution to avoid unnecessary conservatism in design analyses. The Westinghouse Advanced Nodal Code (ANC) incorporates the capability to generate the intra-assembly pin burnup distribution using an efficient algorithm

  7. Verification test of advanced LWR fuel components of Westinghouse type nuclear power plants

    International Nuclear Information System (INIS)

    Kim, Hyung Kyu; Yoon, Kyung Ho; Lee, Young Ho

    2004-08-01

    The purpose of this project is to independently conduct the performance test of the spacer grids and the cladding material of the 16x16 and 17x17 advanced fuels for Westinghouse type plants, and to improve the relevant test technology. Major works and results of the present research are as follows. 1. The design and structural features of the spacer grids were investigated, especially the finally determined I-spring was thoroughly analyzed in the point of the mechanical damage and characteristic. 2. As for the mechanical tests of the space grids, the characterization, the impact and the fretting wear tests were carried out. The block as well as the in-grid tests were conducted for the spring/dimple characterization, from which a simple method was developed that simulated the boundary conditions of the assembled grid straps. The impact tester was modified and improved to accommodate a full size grid assembly. The impact result showed that the grid assembly fulfilled the design criteria. As for the fretting wear tests, a sliding test under the room temperature air/water, a sliding/impact test under the room temperature air and a sliding/impact tests under the high temperature and pressure environments were carried out. To this end, a high temperature and pressure fretting wear tester was newly developed. The wear characteristic and the resistibility of the advanced grid spring/dimple were analyzed in detail. The test results were verified through comparing those with the test results by the Westinghouse company. 3. The properties and performance of the newly adopted material for the cladding, Low Sn Zirlo was investigated by a room and high temperature tensile tests and a corrosion tests under the environments of 360 .deg. C water, 400 steam and 360 .deg. C 70ppm LiOH. Through the present project, all the test equipment and technologies for the fuel components were procured, which will be used for future domestic development of a new fuel

  8. Feasibility analysis of AP1000 wireless communication system and selection of technical solutions

    International Nuclear Information System (INIS)

    Zhao Xin

    2012-01-01

    This article expatiates the rationality and feasibility of AP1000 nuclear power plant adopts wireless communication system as the first choice in routine and emergency operations, compares and analysed. 5 major wireless communication technology solutions, and introduces the Wi-Fi based wireless communication system architecture. (author)

  9. AP1000{sup R} severe accident features and post-Fukushima considerations

    Energy Technology Data Exchange (ETDEWEB)

    Scobel, J. H.; Schulz, T. L.; Williams, M. G. [Westinghouse Electric Company, LLC, 1000 Westinghouse Dr., Cranberry Township, PA 16066 (United States)

    2012-07-01

    The AP1000{sup R} passive nuclear power plant is uniquely equipped to withstand an extended station blackout scenario such as the events following the earthquake and tsunami at Fukushima without compromising core and containment integrity. The AP1000 plant shuts down the reactor, cools the core, containment and spent fuel pool for more than 3 days using passive systems that do not require AC or DC power or operator actions. Following this passive coping period, minimal operator actions are needed to extend the operation of the passive features to 7 days using installed equipment. To provide defense-in-depth for design extension conditions, the AP1000 plant has engineered features that mitigate the effects of core damage. Engineered features retain damaged core debris within the reactor vessel as a key feature. Other aspects of the design protect containment integrity during severe accidents, including unique features of the AP1000 design relative to passive containment cooling with water and air, and hydrogen management. (authors)

  10. AP1000R design robustness against extreme external events - Seismic, flooding, and aircraft crash

    International Nuclear Information System (INIS)

    Pfister, A.; Goossen, C.; Coogler, K.; Gorgemans, J.

    2012-01-01

    Both the International Atomic Energy Agency (IAEA) and the U.S. Nuclear Regulatory Commission (NRC) require existing and new nuclear power plants to conduct plant assessments to demonstrate the unit's ability to withstand external hazards. The events that occurred at the Fukushima-Dai-ichi nuclear power station demonstrated the importance of designing a nuclear power plant with the ability to protect the plant against extreme external hazards. The innovative design of the AP1000 R nuclear power plant provides unparalleled protection against catastrophic external events which can lead to extensive infrastructure damage and place the plant in an extended abnormal situation. The AP1000 plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. The plant's compact safety related footprint and protection provided by its robust nuclear island structures prevent significant damage to systems, structures, and components required to safely shutdown the plant and maintain core and spent fuel pool cooling and containment integrity following extreme external events. The AP1000 nuclear power plant has been extensively analyzed and reviewed to demonstrate that it's nuclear island design and plant layout provide protection against both design basis and extreme beyond design basis external hazards such as extreme seismic events, external flooding that exceeds the maximum probable flood limit, and malicious aircraft impact. The AP1000 nuclear power plant uses fail safe passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems (such as AC power, component cooling water, service water, compressed air or HVAC). The plant has been designed to protect systems, structures, and components critical to placing the reactor in a safe shutdown condition within the steel containment vessel which is

  11. AP1000{sup R} nuclear power plant safety overview for spent fuel cooling

    Energy Technology Data Exchange (ETDEWEB)

    Gorgemans, J.; Mulhollem, L.; Glavin, J.; Pfister, A.; Conway, L.; Schulz, T.; Oriani, L.; Cummins, E.; Winters, J. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    The AP1000{sup R} plant is an 1100-MWe class pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and costs. The AP1000 design uses passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems such as AC power, component cooling water, service water or HVAC. Furthermore, these passive features 'fail safe' during a non-LOCA event such that DC power and instrumentation are not required. The AP1000 also has simple, active, defense-in-depth systems to support normal plant operations. These active systems provide the first level of defense against more probable events and they provide investment protection, reduce the demands on the passive features and support the probabilistic risk assessment. The AP1000 passive safety approach allows the plant to achieve and maintain safe shutdown in case of an accident for 72 hours without operator action, meeting the expectations provided in the U.S. Utility Requirement Document and the European Utility Requirements for passive plants. Limited operator actions are required to maintain safe conditions in the spent fuel pool via passive means. In line with the AP1000 approach to safety described above, the AP1000 plant design features multiple, diverse lines of defense to ensure spent fuel cooling can be maintained for design-basis events and beyond design-basis accidents. During normal and abnormal conditions, defense-in-depth and other systems provide highly reliable spent fuel pool cooling. They rely on off-site AC power or the on-site standby diesel generators. For unlikely design basis events with an extended loss of AC power (i.e., station blackout) or loss of heat sink or both, spent fuel cooling can still be provided indefinitely: - Passive systems, requiring minimal or no operator actions, are sufficient for at least 72 hours under all

  12. AP1000{sup R} design robustness against extreme external events - Seismic, flooding, and aircraft crash

    Energy Technology Data Exchange (ETDEWEB)

    Pfister, A.; Goossen, C.; Coogler, K.; Gorgemans, J. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    Both the International Atomic Energy Agency (IAEA) and the U.S. Nuclear Regulatory Commission (NRC) require existing and new nuclear power plants to conduct plant assessments to demonstrate the unit's ability to withstand external hazards. The events that occurred at the Fukushima-Dai-ichi nuclear power station demonstrated the importance of designing a nuclear power plant with the ability to protect the plant against extreme external hazards. The innovative design of the AP1000{sup R} nuclear power plant provides unparalleled protection against catastrophic external events which can lead to extensive infrastructure damage and place the plant in an extended abnormal situation. The AP1000 plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance and safety. The plant's compact safety related footprint and protection provided by its robust nuclear island structures prevent significant damage to systems, structures, and components required to safely shutdown the plant and maintain core and spent fuel pool cooling and containment integrity following extreme external events. The AP1000 nuclear power plant has been extensively analyzed and reviewed to demonstrate that it's nuclear island design and plant layout provide protection against both design basis and extreme beyond design basis external hazards such as extreme seismic events, external flooding that exceeds the maximum probable flood limit, and malicious aircraft impact. The AP1000 nuclear power plant uses fail safe passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems (such as AC power, component cooling water, service water, compressed air or HVAC). The plant has been designed to protect systems, structures, and components critical to placing the reactor in a safe shutdown condition within the steel

  13. Best estimate probabilistic safety assessment results for the Westinghouse Advanced Loop Tester (WALT)

    International Nuclear Information System (INIS)

    Wang, Guoqiang; Xu, Yiban; Oelrich, Robert L. Jr.; Byers, William A.; Young, Michael Y.; Karoutas, Zeses E.

    2011-01-01

    The nuclear industry uses the probabilistic safety assessment (PSA) technique to improve safety decision making and operation. The methodology evaluates the system reliability, which is defined as the probability of system success, and the postulated accident/problematic scenarios of systems for the nuclear power plants or other facilities. The best estimate probabilistic safety assessment (BE-PSA) method of evaluating system reliability and postulated problematic scenarios will produce more detailed results of interest, such as best estimated reliability analysis and detailed thermal hydraulic calculations using a sub-channel or Computational Fluid Dynamics (CFD) code. The methodology is typically applied to reactors, but can also be applied to any system such as a test facility. In this paper, a BE-PSA method is introduced and used for evaluating the Westinghouse Advanced Loop Tester (WALT). The WALT test loop at the George Westinghouse Science and Technology Center (STC), which was completed in October 2005, is designed to be utilized to model the top grid span of a hot rod in a fuel assembly under the Pressurizer Water Reactor (PWR) normal operating conditions. In order to safely and successfully operate the WALT test loop and correctly use the WALT experimental data, it is beneficial to perform a probabilistic safety assessment and analyze the thermal hydraulic results for the WALT loop in detail. Since October 2005, a number of test runs have been performed on the WALT test facility designed and fabricated by Westinghouse Electric Company LLC. This paper briefly describes the BE-PSA method and performs BE-PSA for the WALT loop. Event trees linked with fault trees embedding thermal hydraulic analysis models, such as sub-channel and/or CFD models, were utilized in the analyses. Consequently, some selected useful experimental data and analysis results are presented for future guidance on WALT and/or other similar test facilities. For example, finding and

  14. Performance Analysis of AP1000 Passive Systems during Direct Vessel Injection (DVI Line Break

    Directory of Open Access Journals (Sweden)

    A.S. Ekariansyah

    2016-08-01

    Full Text Available Generation II Nuclear Power Plants (NPPs have a design weakness as shown by the Fukushima accident. Therefore, Generation III+ NPPs are developed with focus on improvements of fuel technology and thermal efficiency, standardized design, and the use of passive safety system. One type of Generation III+ NPP is the AP1000 that is a pressurized water reactor (PWR type that has received the final design acceptance from US-NRC and is already under construction at several sites in China as of 2015. The aim of this study is to investigate the behavior and performance of the passive safety system in the AP1000 and to verify the safety margin during the direct vessel injection (DVI line break as selected event. This event was simulated using RELAP5/SCDAP/Mod3.4 as a best-estimate code developed for transient simulation of light water reactors during postulated accidents. This event is also described in the AP1000 design control document as one of several postulated accidents simulated using the NOTRUMP code. The results obtained from RELAP5 calculation was then compared with the results of simulations using the NOTRUMP code. The results show relatively good agreements in terms of time sequences and characteristics of some injected flow from the passive safety system. The simulation results show that the break of one of the two available DVI lines can be mitigated by the injected coolant flowing, which is operated effectively by gravity and density difference in the cooling system and does not lead to core uncovery. Despite the substantial effort to obtain an apropriate AP1000 model due to lack of detailed geometrical data, the present model can be used as a platform model for other initiating event considered in the AP1000 accident analysis.

  15. 77 FR 74696 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on AP-1000...

    Science.gov (United States)

    2012-12-17

    ... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on AP-1000; Notice of Meeting The ACRS Subcommittee on AP-1000 will hold a meeting on January 18, 2013, Room T-2B1, 11545 Rockville Pike, Rockville, Maryland. The entire meeting will be open to public...

  16. Reactor coolant system hydrostatic test and risk analysis for the first AP1000 unit

    International Nuclear Information System (INIS)

    Cao Hongjun; Yan Xiuping

    2013-01-01

    The cold hydrostatic test scheme of the primary coolant circuit, of the first AP1000 unit was described. Based on the up-stream design documents, standard specifications and design technical requirements, the select principle of test boundary was identified. The design requirements for water quality, pressure, temperature and temporary hydro-test pump were proposed. A reasonable argument for heating and pressurization rate, and cooling and depressurization rate was proposed. The possible problems and risks during the hydrostatic test were analyzed. This test scheme can provide guidance for the revisions and implementations of the follow-up test procedures. It is a good reference for hydrostatic tests of AP1000 units in the future in China. (authors)

  17. Study and practice on NI design management of the first AP1000 plant

    International Nuclear Information System (INIS)

    Jiang Feizhou

    2014-01-01

    For difficulties on NI design management during the construction of the world's first AP1000 unit, Sanmen Nuclear Power Corporation has established the effective method to improve the management of the NI design and the project construction based on the practice and improvement. This paper introduces difficulties on NI design management, and analyzes the causes of problems, and put forward recommended measures, to provide help and reference for the construction of similar nuclear power plants. (authors)

  18. Refer to AP1000 for discussing the betterment of seismic design of internal nuclear power plant

    International Nuclear Information System (INIS)

    Gong Zhenbang; Zhang Renyan

    2014-01-01

    As a reference technique of AP1000, This paper discussed the betterment of seismic design of nuclear power plant in three ways. (1) Establish design criteria and guidelines for protection from seismic interaction; (2) Nuclear power plant seismic design of eliminating or weaken operation-basis earthquake; (3) Develop the seismic margin analysis (SMA) of the nuclear power plant. These three aspect are frontier technology in internal seismic design of internal nuclear power plant, and also these three technology are related intimately. (authors)

  19. Spent fuel pool spray cooling system for the AP1000 {sup registered}

    Energy Technology Data Exchange (ETDEWEB)

    Vujic, Zoran; Sassen, Felix; Tietsch, Wolfgang [Westinghouse Electric Germany GmbH, Mannheim (Germany)

    2013-07-01

    The AP1000 {sup registered} plant design features multiple, diverse lines of defense to ensure spent fuel cooling can be maintained for Design Basis Events and Beyond Design Basis Accidents (BDBA). The AP1000 {sup registered} plant lines of defense with respect to Spent Fuel Pool (SFP) cooling are as follows: 1. During normal and abnormal conditions, defense-in-depth and duty systems provide highly reliable SFP cooling, supplied by offsite AC power or the onsite Standby Diesel Generators. 2. For unlikely events with extended loss of AC power (i.e. station black-out) and/or loss of heat sink, spent fuel cooling can be still provided indefinitely by: 2a. Passive systems, requiring minimal or no operator actions, sufficient for at least 72 hours under all possible loading conditions. 2b. After 3 days, several different means are provided to continue SFP cooling using installed plant equipment as well as off-site equipment with built-in connections. 3. Even for BDBA with postulated SFP damage and multiple failures in the passive safety-related systems and in the defense-in-depth active systems, the AP1000 {sup registered} SFP Spray System provides an additional line of defense to prevent spent fuel damage. (orig.)

  20. Effect of air condition on AP-1000 containment cooling performance in station black out accident

    International Nuclear Information System (INIS)

    Hendro Tjahjono

    2015-01-01

    AP1000 reactor is a nuclear power plant generation III+ 1000 MWe which apply passive cooling concept to anticipate accidents triggered by the extinction of the entire supply of electrical power or Station Black Out (SBO). In the AP1000 reactor, decay heat disposal mechanism conducted passively through the PRHR-IRWST and subsequently forwarded to the reactor containment. Containment externally cooled through natural convection in the air gap and through evaporation cooling water poured on the outer surface of the containment wall. The mechanism of evaporation of water into the air outside is strongly influenced by the conditions of humidity and air temperature. The purpose of this study was to determine the extent of the influence of the air condition on cooling capabilities of the AP1000 containment. The method used is to perform simulations using Matlab-based analytical calculation model capable of estimating the power of heat transferred. The simulation results showed a decrease in power up to 5% for relative humidity rose from 10% to 95%, while the variation of air temperature of 10°C to 40°C, the power will decrease up to 15%. It can be concluded that the effect of air temperature increase is much more significant in lowering the containment cooling ability compared with the increase of humidity. (author)

  1. The Design of Cooling System Model on The AP1000 Containment

    International Nuclear Information System (INIS)

    Daddy Setyawan; Yerri Noer Kartiko; Aryadi Suwono; Ari Darmawan Pasek; Nathanael P Tandian; Efrizon Umar

    2009-01-01

    The policy of national energy leads to the utilization of new energy as nuclear energy, and also contains some efforts to increase reactor safety and optimizing in the design of safety system component such as passive cooling system on reactor containment tank. Because of this, the assessment of safety level to passive safety system needs to be made. To increase the understanding it, the design of cooling system model on containment tank should be done to get safety level on cooling system in the AP1000 containment. To reach the similar model with reality and inexpensive cost, we should make assessment about similarity and dimensionless number. While the heat transfer of air natural circulation and water spray cooling system are a result of gravity approach, we can calculate Grashof modification number and Reynolds number respectively. By this approach, we have a factor of forty for laboratory model. From this model, we hope that we get characteristic correlation to heat transfer on the containment of AP1000 for both air natural circulation and water spray result from gravity. Finally, we can assess the safety level of passive cooling system on the AP1000 containment. (author)

  2. AP1000 station blackout study with and without depressurization using RELAP5/SCDAPSIM

    Energy Technology Data Exchange (ETDEWEB)

    Trivedi, A.K. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Allison, C. [Innovative Systems Software Idaho Falls, ID 83406 (United States); Khanna, A., E-mail: akhanna@iitk.ac.in [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India); Munshi, P. [Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016 (India)

    2016-10-15

    Highlights: • A representative RELAP5/SCDAPSIM model of AP1000 has been developed. • Core is modeled using SCDAP. • A SBO for the AP1000 has been simulated for high pressure (no depressurization) and low pressure (depressurization). • Significant differences in the damage progression have been observed for the two cases. • Results also reinforced the fact that surge line fails before vessel failure in case of high pressure scenario. - Abstract: Severe accidents like TMI-2, Chernobyl, Fukushima made it inevitable to analyze station blackout (SBO) for all the old as well as new designs although it is not a regulatory requirement in most of the countries. For such improbable accidents, a SBO for the AP1000 using RELAP5/SCDAPSIM has been simulated. Many improvements have been made in fuel damage progression models of SCDAP after the Fukushima accident which are now being tested for the new reactor designs. AP1000 is a 2-loop pressurized water reactor (PWR) with all the emergency core cooling systems based on natural circulation. Its core design is very similar to 3-loop PWR with 157 fuel assemblies. The primary circuit pumps, pressurizer and steam generators (with necessary secondary side) are modeled using RELAP5. The core has been divided into 20 axial nodes and 6 radial rings; the corresponding six groups of assemblies have been modeled as six pipe components with proportionate flow area. Fuel assemblies are modeled using SCDAP fuel and control components. SCDAP has 2d-heat conduction and radiative heat transfer, oxidation and complete severe fuel damage progression models. The final input deck achieved all the steady state thermal hydraulic conditions comparable to the design control document of AP1000. To quantify the core behavior, under unavailability of all safety systems, various time profiles for SBO simulations @ high pressure and low pressure have been compared. This analysis has been performed for 102% (3468 MWt) of the rated core power. The

  3. Plan for fully decontaminating and decommissioning of the Westinghouse Advanced Reactors Division Fuel Laboratories at Cheswick, Revision 3

    International Nuclear Information System (INIS)

    1982-01-01

    The project scope of work included the complete decontamination and decommissioning (D and D) of the Westinghouse ARD Fuel Laboratories at the Cheswick Site in the shortest possible time. This has been accomplished in the following four phases: (1) preparation of documents and necessary paperwork; packaging and shipping of all special nuclear materials in an acceptable form to a reprocessing agency; (2) decontamination of all facilities, glove boxes and equipment; loading of generated waste into bins, barrels and strong wooden boxes; (3) shipping of all bins, barrels and boxes containing waste to the designated burial site; removal of all utility services from the laboratories; (4) final survey of remaining facilities and certification for nonrestricted use; preparation of final report. This volume contains the following 3 attachments: (1) Plan for Fully Decontamination and Decommissioning of the Westinghouse Advanced Reactors Division Fuel Laboratories at Cheswick; (2) Environmental Assessment for Decontamination and Decommissioning the Westinghouse Advanced Reactors Division Plutonium Fuel Laboratories, Cheswick, PA; and (3) WARD-386, Quality Assurance Program Description for Decontamination and Decommissioning Activities

  4. Radioactive waste shipments to Hanford retrievable storage from Westinghouse Advanced Reactors and Nuclear Fuels Divisions, Cheswick, Pennsylvania

    International Nuclear Information System (INIS)

    Duncan, D.; Pottmeyer, J.A.; Weyns, M.I.; Dicenso, K.D.; DeLorenzo, D.S.

    1994-04-01

    During the next two decades the transuranic (TRU) waste now stored in the burial trenches and storage facilities at the Hanford Sits in southeastern Washington State is to be retrieved, processed at the Waste Receiving and Processing Facility, and shipped to the Waste Isolation Pilot Plant (WIPP), near Carlsbad, New Mexico for final disposal. Approximately 5.7 percent of the TRU waste to be retrieved for shipment to WIPP was generated by the decontamination and decommissioning (D ampersand D) of the Westinghouse Advanced Reactors Division (WARD) and the Westinghouse Nuclear Fuels Division (WNFD) in Cheswick, Pennsylvania and shipped to the Hanford Sits for storage. This report characterizes these radioactive solid wastes using process knowledge, existing records, and oral history interviews

  5. ANALISIS SKENARIO KEGAGALAN SISTEM UNTUK MENENTUKAN PROBABILITAS KECELAKAAN PARAH AP1000

    Directory of Open Access Journals (Sweden)

    D.T. Sony Tjahyani

    2014-03-01

    Full Text Available Kejadian Fukushima telah menunjukkan bahwa kecelakaan parah dapat terjadi, maka dari itu sangatlah penting untuk menganalisis tingkat keselamatan pada reaktor daya. Berdasarkan rekomendasi expert mission IAEA setelah kejadian Fukushima, perlu dilakukan upaya untuk meminimalisasi terjadinya kecelakaan parah yaitu dengan melakukan proses pendinginan yang maksimal. Dalam konsep keselamatan fasilitas nuklir, khususnya reaktor daya telah diterapkan konsep keselamatan berlapis (Defence in Depth, DiD. Konsep keselamatan tersebut terdiri atas 5 level pertahanan yang bertujuan mencegah dan mengurangi lepasan produk fisi ke masyarakat dan lingkungan pada saat reaktor daya mengalami kecelakaan. Dalam reaktor telah didesain sistem atau tindakan yang mempunyai fungsi untuk mengatasi setiap level tersebut. Tujuan dari analisis ini adalah menentukan probabilitas kecelakaan parah dengan melakukan skenario kegagalan sistem dalam proses pendinginan di reaktor. Sebagai obyek analisis adalah reaktor daya AP1000, karena jenis reaktor ini sedang banyak dibangun saat ini. Skenario dilakukan dengan mengasumsikan beberapa kombinasi kegagalan sistem yang termasuk dalam DiD level 2 dan 3. Kegagalan sistem kemudian dianalisis dengan menggunakan analisis pohon kegagalan berdasarkan perangkat lunak SAPHIRE ver. 6.76. Dari analisis didapatkan probabilitas gagal dari kelompok sistem DiD level 2 dan 3 pada AP1000 masih di bawah batas kriteria dari IAEA yaitu lebih kecil dari 10-2, serta probabilitas kecelakaan parah didapatkan sebesar 6,17 x 10-10. Berdasarkan analisis ini disimpulkan bahwa AP1000 mempunyai tingkat keselamatan yang cukup tinggi, karena melalui skenario kegagalan sistem didapatkan probabilitas kecelakaan parah yang sangat kecil.   ABSTRACT Fukushima accident has shown that severe accident could be occurred, therefore it is important to analyze safety level of nuclear power plants. Based on the recommendations of IAEA expert mission after the Fukushima accident

  6. The AP1000R nuclear power plant innovative features for extended station blackout mitigation

    International Nuclear Information System (INIS)

    Vereb, F.; Winters, J.; Schulz, T.; Cummins, E.; Oriani, L.

    2012-01-01

    Station Blackout (SBO) is defined as 'a condition wherein a nuclear power plant sustains a loss of all offsite electric power system concurrent with turbine trip and unavailability of all onsite emergency alternating current (AC) power system. Station blackout does not include the loss of available AC power to buses fed by station batteries through inverters or by alternate AC sources as defined in this section, nor does it assume a concurrent single failure or design basis accident...' in accordance with Reference 1. In this paper, the innovative features of the AP1000 plant design are described with their operation in the scenario of an extended station blackout event. General operation of the passive safety systems are described as well as the unique features which allow the AP1000 plant to cope for at least 7 days during station blackout. Points of emphasis will include: - Passive safety system operation during SBO - 'Fail-safe' nature of key passive safety system valves; automatically places the valve in a conservatively safe alignment even in case of multiple failures in all power supply systems, including normal AC and battery backup - Passive Spent Fuel Pool cooling and makeup water supply during SBO - Robustness of AP1000 plant due to the location of key systems, structures and components required for Safe Shutdown - Diverse means of supplying makeup water to the Passive Containment Cooling System (PCS) and the Spent Fuel Pool (SFP) through use of an engineered, safety-related piping interface and portable equipment, as well as with permanently installed onsite ancillary equipment. (authors)

  7. The AP1000{sup R} nuclear power plant innovative features for extended station blackout mitigation

    Energy Technology Data Exchange (ETDEWEB)

    Vereb, F.; Winters, J.; Schulz, T.; Cummins, E.; Oriani, L. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    Station Blackout (SBO) is defined as 'a condition wherein a nuclear power plant sustains a loss of all offsite electric power system concurrent with turbine trip and unavailability of all onsite emergency alternating current (AC) power system. Station blackout does not include the loss of available AC power to buses fed by station batteries through inverters or by alternate AC sources as defined in this section, nor does it assume a concurrent single failure or design basis accident...' in accordance with Reference 1. In this paper, the innovative features of the AP1000 plant design are described with their operation in the scenario of an extended station blackout event. General operation of the passive safety systems are described as well as the unique features which allow the AP1000 plant to cope for at least 7 days during station blackout. Points of emphasis will include: - Passive safety system operation during SBO - 'Fail-safe' nature of key passive safety system valves; automatically places the valve in a conservatively safe alignment even in case of multiple failures in all power supply systems, including normal AC and battery backup - Passive Spent Fuel Pool cooling and makeup water supply during SBO - Robustness of AP1000 plant due to the location of key systems, structures and components required for Safe Shutdown - Diverse means of supplying makeup water to the Passive Containment Cooling System (PCS) and the Spent Fuel Pool (SFP) through use of an engineered, safety-related piping interface and portable equipment, as well as with permanently installed onsite ancillary equipment. (authors)

  8. Development of Fuel ROd Behavior Analysis code (FROBA) and its application to AP1000

    International Nuclear Information System (INIS)

    Yu, Hongxing; Tian, Wenxi; Yang, Zhen; SU, G.H.; Qiu, Suizheng

    2012-01-01

    Highlights: ► A Fuel ROd Behavior Analysis code (FROBA) has been developed. ► The effects irradiation and burnup has been considered in FROBA. ► The comparison with INL’s results shows a good agreement. ► The FROBA code was applied to AP1000. ► Peak fuel temperature, gap width, hoop strain, etc. were obtained. -- Abstract: The reliable prediction of nuclear fuel rod behavior is of great importance for safety evaluation of nuclear reactors. In the present study, a thermo-mechanical coupling code FROBA (Fuel ROd Behavior Analysis) has been independently developed with consideration of irradiation and burnup effects. The thermodynamic, geometrical and mechanical behaviors have been predicted and were compared with the results obtained by Idaho National Laboratory to validate the reliability and accuracy of the FROBA code. The validated code was applied to analyze the fuel behavior of AP1000 at different burnup levels. The thermal results show that the predicted peak fuel temperature experiences three stages in the fuel lifetime. The mechanical results indicate that hoop strain at high power is greater than that at low power, which means that gap closure phenomenon will occur earlier at high power rates. The maximum cladding stress meets the requirement of yield strength limitation in the entire fuel lifetime. All results show that there are enough safety margins for fuel rod behavior of AP1000 at rated operation conditions. The FROBA code is expected to be applied to deal with more complicated fuel rod scenarios after some modifications.

  9. Shadow management applied in the first AP1000 project under the islands contract condition

    International Nuclear Information System (INIS)

    Liu Xiao

    2010-01-01

    As the global first AP1000 nuclear project, Sanmen phase I nuclear project itself has many challenges from design, procurement to construction managements for non practical nuclear project and experience can be referenced. Islands contract pattern was adopted by this project and this contract pattern has its own strength and weakness. Considering the negative influence result from the first unit, this project has the great postpone risk. Shadow management here tries to reduce these risks and enhance the project surveillance and control by the owner to promote the final goal of this project. (authors)

  10. Reliability analysis on passive residual heat removal of AP1000 based on Grey model

    Energy Technology Data Exchange (ETDEWEB)

    Qi, Shi; Zhou, Tao; Shahzad, Muhammad Ali; Li, Yu [North China Electric Power Univ., Beijing (China). School of Nuclear Science and Engineering; Beijing Key Laboratory of Passive Safety Technology for Nuclear Energy, Beijing (China); Jiang, Guangming [Nuclear Power Institute of China, Chengdu (China). Science and Technology on Reactor System Design Technology Laboratory

    2017-06-15

    It is common to base the design of passive systems on the natural laws of physics, such as gravity, heat conduction, inertia. For AP1000, a generation-III reactor, such systems have an inherent safety associated with them due to the simplicity of their structures. However, there is a fairly large amount of uncertainty in the operating conditions of these passive safety systems. In some cases, a small deviation in the design or operating conditions can affect the function of the system. The reliability of the passive residual heat removal is analysed.

  11. Grain size control method for the nozzles of AP1000 primary coolant pipes

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Shenglong [State Key Laboratory for Advanced Metals and Materials, University of Science & Technology Beijing, Beijing 100083 (China); Sun, Yanhui [Collaborative Innovation Center of Steel Technology, University of Science & Technology Beijing, Beijing 100083 (China); Yang, Bin, E-mail: byang@ustb.edu.cn [State Key Laboratory for Advanced Metals and Materials, University of Science & Technology Beijing, Beijing 100083 (China); Collaborative Innovation Center of Steel Technology, University of Science & Technology Beijing, Beijing 100083 (China); Zhang, Mingxian [State Key Laboratory for Advanced Metals and Materials, University of Science & Technology Beijing, Beijing 100083 (China)

    2017-04-01

    Highlights: • Design a new forging technology for AP1000 primary coolant pipe. • Method combining FEM and scale-down experiments is adopted. • The grain size and distribution in simulation and experiment are consistent. • Get optimal forging parameters for production guiding. - Abstract: AP1000 primary coolant pipe is made of 316LN austenitic stainless steel. It is a large special-shaped pipe manufactured by integral forging technology. Owing to non-uniform temperature and deformation during forging, coarse grains often occur in the boss sections of the pipe especially in the nozzles’ parts. In the present study, a new forging technology was proposed to control the grain size. The finite element method was used to optimize the forging speed and friction coefficient, then the scale-down experiments were performed for comparison. The forging speed is suggested to be less than 20 mm/s, and effective lubricants should be used to decrease the friction coefficient. The errors of the grain size between the experiment and simulation are less than 20%.

  12. Grain size control method for the nozzles of AP1000 primary coolant pipes

    International Nuclear Information System (INIS)

    Wang, Shenglong; Sun, Yanhui; Yang, Bin; Zhang, Mingxian

    2017-01-01

    Highlights: • Design a new forging technology for AP1000 primary coolant pipe. • Method combining FEM and scale-down experiments is adopted. • The grain size and distribution in simulation and experiment are consistent. • Get optimal forging parameters for production guiding. - Abstract: AP1000 primary coolant pipe is made of 316LN austenitic stainless steel. It is a large special-shaped pipe manufactured by integral forging technology. Owing to non-uniform temperature and deformation during forging, coarse grains often occur in the boss sections of the pipe especially in the nozzles’ parts. In the present study, a new forging technology was proposed to control the grain size. The finite element method was used to optimize the forging speed and friction coefficient, then the scale-down experiments were performed for comparison. The forging speed is suggested to be less than 20 mm/s, and effective lubricants should be used to decrease the friction coefficient. The errors of the grain size between the experiment and simulation are less than 20%.

  13. Post-accident cooling capacity analysis of the AP1000 passive spent fuel pool cooling system

    International Nuclear Information System (INIS)

    Su Xia

    2013-01-01

    The passive design is used in AP1000 spent fuel pool cooling system. The decay heat of the spent fuel is removed by heating-boiling method, and makeup water is provided passively and continuously to ensure the safety of the spent fuel. Based on the analysis of the post-accident cooling capacity of the spent fuel cooling system, it is found that post-accident first 72-hour cooling under normal refueling condition and emergency full-core offload condition can be maintained by passive makeup from safety water source; 56 hours have to be waited under full core refueling condition to ensure the safety of the core and the spent fuel pool. Long-term cooling could be conducted through reserved safety interface. Makeup measure is available after accident and limited operation is needed. Makeup under control could maintain the spent fuel at sub-critical condition. Compared with traditional spent fuel pool cooling system design, the AP1000 design respond more effectively to LOCA accidents. (authors)

  14. Experimental research of liquid entrainment through ADS-4 in AP1000

    International Nuclear Information System (INIS)

    Meng, Zhaoming; Dong, Bo; Wang, Laishun; Fu, Xiaoliang; Tian, Wenxi; Yang, Yanhua; Su, Guanghui

    2014-01-01

    Highlights: • We performed experimental research of liquid entrainment through ADS-4 in AP1000. • Effect of various factors on entrainment at T-junction was conducted. • Visualization research was conducted to make entrainment mechanism clear. - Abstract: In this study, based on a T-junction that consists of Automatic Depressurization System Stage Four (ADS-4) and hot leg in an AP1000 plant, a small-scale experimental research on entrainment at a T-junction was performed. This study mainly focused on the effect of various factors on entrainment, such as the effect of branch size, branch shape and liquid crossflow. The flow pattern map was plotted from the experimental data, and the visualization research indicated that the entrainment phenomena through a large size branch were apparently different from that through a small branch. Three entrainment phenomena were observed in the studies, two entrainment mechanisms could be found in the stratified flow regime entrainment area, the existence of branch contributed to generating intermittent flow in the horizontal main pipe, and the backflow region was observable in the vicinity of a large size branch inlet. Also, experimental research showed that downstream of the branch of T-junction had an important effect on the onset entrainment, and liquid crossflow did not seem to affect the onset entrainment

  15. Numerical simulation of AP1000 LBLOCA with SCDAP/RELAP 4.0 code

    International Nuclear Information System (INIS)

    Xie Heng

    2017-01-01

    The risk of large-break loss of coolant accident (LBLOCA) is that core will be exposed once the accident occurs, and may cause core damages. New phenomena may occur in LBLOCA due to passive safety injection adopted by AP1000. This paper used SCDAP/RELAP5 4.0 to build the numerical model of AP1000 and double-end guillotine of cold leg is simulated. Reactor coolant system and passive core cooling system were modeled by RELAP5 modular. HEAT STRUCTURE component of RELAP5 was used to simulate the fuel rod. The reflood option in RELAP5 was chosen to be activated or not to study the effect of axial heat conduction. Results show that the axial heat conduction plays an important role in the reflooding phase and can effectively shorten reflood process. An alternative core model is built by SCDAP modular. It is found that the SCDAP model predicts higher maximum peak cladding temperature and longer reflood process than RELAP5 model. Analysis shows that clad oxidation heat plays a key role in the reflood. From the simulation results, it can be concluded that the cladding will keep intact and fission product will not be released from fuel to coolant in LBLOCA. (author)

  16. Westinghouse Small Modular Reactor (SMR) Programe

    International Nuclear Information System (INIS)

    Shulyak, Nick

    2014-01-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) in which all primarycomponents associated with the nuclear steam supply system, including the steam generator and the pressurizer, are housed within the reactor vessel. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. This paper describes the design and functionality of the Westinghouse SMR, the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design drivers include safety, economics, reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR safety system design is passive, is based largely on the passive safety systems used in the AP1000 reactor, and provides mitigation of all design basis accidents without the need for offsite AC electrical power for a period of seven days. The economics of the Westinghouse SMR challenges the established approach of large Light Water Reactors (LWR) that utilized the economies of scale to reach economic competiveness. To serve the market expectation of smaller capital investment and cost competitive energy, a modular design approach is implemented within the Westinghouse SMR. The Westinghouse SMR building layout integrates the three basic design constraints of modularization; transportation, handling and module-joining technology. The integral Westinghouse SMR design eliminates large loop piping, which significantly reduces the flow area of postulated loss of coolant accidents (LOCAs). The Westinghouse SMR containment is a high

  17. Analysis for low and intermediate level radioactive waste disposal of AP1000 nuclear power in China

    International Nuclear Information System (INIS)

    Yang Bin; Ren Li; Hua Wei; Ma Xiaoqiang; Ma Ruoxia; Fang Xianghong

    2014-01-01

    AP1000 nuclear power is adopted in Sanmen Nuclear Power Plant, Zhejiang province and Haiyang Nuclear Power Plant, Shandong province. The filter and resin generated by the operation of nuclear power plants are handled by cement cured process and compression technology in Sanmen Nuclear Power Plant, while they are loaded in polyethylene HIC in Haiyang Nuclear Power Plant. At present, there were not engineering practice for dispose polyethylene HIC in China. The document discusses three disposal options of polyethylene HIC: one is that polyethylene HIC placed in over pack, another is that polyethylene HIC mixed with other drum, the third is that polyethylene HIC placed in disposal unit which is divided into small units. (authors)

  18. Study on radioactive release of gaseous and liquid effluents during normal operation of AP1000

    International Nuclear Information System (INIS)

    Gong Quan; Zhou Jing; Liu Yu

    2014-01-01

    The gaseous and liquid radioactive releases of pressurized water reactors plant during normal operation are an important content of environmental impact assessment and play a significant role in the design of nuclear power plant. According to the design characters of AP1OOO radioactive waste management system and the study on the calculation method and the release pathways, the calculation model of the gaseous and liquid radioactive releases during normal operation for AP1OOO are established. Base on the established calculation model and the design parameters of AP1000, the expected value of gaseous and liquid radioactive releases of AP1OOO is calculated. The results of calculation are compared with the limits in GB 6249-2011 and explain the adder that is included tu account for anticipated operational occurrences, providing a reference for environmental impact assessment of pressurized water reactor. (authors)

  19. Application of status uncertainty analysis methods for AP1000 LBLOCA calculation

    International Nuclear Information System (INIS)

    Zhang Shunxiang; Liang Guoxing

    2012-01-01

    Parameter uncertainty analysis is developed by using the reasonable method to establish the response relations between input parameter uncertainties and output uncertainties. The application of the parameter uncertainty analysis makes the simulation of plant state more accuracy and improves the plant economy with reasonable security assurance. The AP1000 LBLOCA was analyzed in this paper and the results indicate that the random sampling statistical analysis method, sensitivity analysis numerical method and traditional error propagation analysis method can provide quite large peak cladding temperature (PCT) safety margin, which is much helpful for choosing suitable uncertainty analysis method to improve the plant economy. Additionally, the random sampling statistical analysis method applying mathematical statistics theory makes the largest safety margin due to the reducing of the conservation. Comparing with the traditional conservative bounding parameter analysis method, the random sampling method can provide the PCT margin of 100 K, while the other two methods can only provide 50-60 K. (authors)

  20. Application case study of AP1000 automatic depressurization system (ADS) for reliability evaluation by GO-FLOW methodology

    Energy Technology Data Exchange (ETDEWEB)

    Hashim, Muhammad, E-mail: hashimsajid@yahoo.com; Hidekazu, Yoshikawa, E-mail: yosikawa@kib.biglobe.ne.jp; Takeshi, Matsuoka, E-mail: mats@cc.utsunomiya-u.ac.jp; Ming, Yang, E-mail: myang.heu@gmail.com

    2014-10-15

    Highlights: • Discussion on reasons why AP1000 equipped with ADS system comparatively to PWR. • Clarification of full and partial depressurization of reactor coolant system by ADS system. • Application case study of four stages ADS system for reliability evaluation in LBLOCA. • GO-FLOW tool is capable to evaluate dynamic reliability of passive safety systems. • Calculated ADS reliability result significantly increased dynamic reliability of PXS. - Abstract: AP1000 nuclear power plant (NPP) utilized passive means for the safety systems to ensure its safety in events of transient or severe accidents. One of the unique safety systems of AP1000 to be compared with conventional PWR is the “four stages Automatic Depressurization System (ADS)”, and ADS system originally works as an active safety system. In the present study, authors first discussed the reasons of why four stages ADS system is added in AP1000 plant to be compared with conventional PWR in the aspect of reliability. And then explained the full and partial depressurization of RCS system by four stages ADS in events of transient and loss of coolant accidents (LOCAs). Lastly, the application case study of four stages ADS system of AP1000 has been conducted in the aspect of reliability evaluation of ADS system under postulated conditions of full RCS depressurization during large break loss of a coolant accident (LBLOCA) in one of the RCS cold legs. In this case study, the reliability evaluation is made by GO-FLOW methodology to determinate the influence of ADS system in dynamic reliability of passive core cooling system (PXS) of AP1000, i.e. what will happen if ADS system fails or successfully actuate. The GO-FLOW is success-oriented reliability analysis tool and is capable to evaluating the systems reliability/unavailability alternatively to Fault Tree Analysis (FTA) and Event Tree Analysis (ETA) tools. Under these specific conditions of LBLOCA, the GO-FLOW calculated reliability results indicated

  1. Effects of the Application of the New Nuclear Data Library ENDF/B to the Criticality Analysis of AP1000

    Science.gov (United States)

    Kuntoro, Iman; Sembiring, T. M.; Susilo, Jati; Deswandri; Sunaryo, G. R.

    2018-02-01

    Calculations of criticality of the AP1000 core due to the use of new edition of nuclear data library namely ENDF/B-VII and ENDF/B-VII.1 have been done. This work is aimed to know the accuracy of ENDF/B-VII.1 compared to ENDF/B-VII and ENDF/B-VI.8. in determining the criticality parameter of AP1000. Analysis ws imposed to core at cold zero power (CZP) conditions. The calculations have been carried out by means of MCNP computer code for 3 dimension geometry. The results show that criticality parameter namely effective multiplication factor of the AP1000 core are higher than that ones resulted from ENDF/B-VI.8 with relative differences of 0.39% for application of ENDF/B-VII and of 0.34% for application of ENDF/B-VII.1.

  2. Reactivity Coefficient Calculation for AP1000 Reactor Using the NODAL3 Code

    Science.gov (United States)

    Pinem, Surian; Malem Sembiring, Tagor; Tukiran; Deswandri; Sunaryo, Geni Rina

    2018-02-01

    The reactivity coefficient is a very important parameter for inherent safety and stability of nuclear reactors operation. To provide the safety analysis of the reactor, the calculation of changes in reactivity caused by temperature is necessary because it is related to the reactor operation. In this paper, the temperature reactivity coefficients of fuel and moderator of the AP1000 core are calculated, as well as the moderator density and boron concentration. All of these coefficients are calculated at the hot full power condition (HFP). All neutron diffusion constant as a function of temperature, water density and boron concentration were generated by the SRAC2006 code. The core calculations for determination of the reactivity coefficient parameter are done by using NODAL3 code. The calculation results show that the fuel temperature, moderator temperature and boron reactivity coefficients are in the range between -2.613 pcm/°C to -4.657pcm/°C, -1.00518 pcm/°C to 1.00649 pcm/°C and -9.11361 pcm/ppm to -8.0751 pcm/ppm, respectively. For the water density reactivity coefficients, the positive reactivity occurs at the water temperature less than 190 °C. The calculation results show that the reactivity coefficients are accurate because the results have a very good agreement with the design value.

  3. Local panels and maintainability human factors assessment for AP1000 nuclear power plant

    International Nuclear Information System (INIS)

    Li, Zhonghai; Reed, Julie I.

    2011-01-01

    A document entitled 'AP1000 Local Panels and Maintainability Human Factors Design Guidelines' was produced to aid the designers to specifically include human factors (HF) considerations in the design, operation, and maintenance of local control stations and plant equipment. To ensure that the applicable HF design guidelines are appropriately applied to the design of local panels and maintenance activities, and identify any HF improvement opportunities that can readily be implemented at the design stage, a HF assessment of maintenance activities and local plant operations is underway. This assessment gives priority to local control stations and equipment which have been identified as having a potential impact on safety. This includes risk-significant systems, structures and components (SSCs) identified through the probabilistic risk assessment (PRA), and local operator actions as required by the Emergency Operating Procedures (EOPs). Local actions, maintenance activities and associated operator interfaces are reviewed against the relevant HF guidelines. The results of the assessment include a description of the component, associated local actions and/or required maintenance activities, good design features and/or potential issues, and recommendations for change or improvement. These results are communicated to responsible design engineers who evaluate the impact to plant design and implement design changes, if deemed necessary. (author)

  4. Research on the improvement design for the attachment of supports to AP1000 module wall

    International Nuclear Information System (INIS)

    Li Cheng; Liu Jianwei; Shan Ying

    2013-01-01

    Background: Modularization is one of the main characteristics for AP1000 nuclear power plant building. The steel-concrete-steel module wall is used instead of reinforced concrete structure wall. Usually, lots of Overlay Plate Embedments will be installed on the module wall to connect and fasten other structures, such as pipes, equipment and operation platforms. As for many supports taking less design loads, the safety margin is too big when using OLP embedment. Purpose: An improvement design will make sense that the supports with less design loads can be welded directly to the module wall instead of embedments. Methods: A finite element analysis based on nuclear-related concrete code is carried out. Results: Through analysis, the equations for the allowable design loads of supports to be welded directly to module wall are provided in this paper. Conclusions: The improvement design is proved feasible. In this way, the strength for steel face plate and studs will be utilized fully and this method will facilitate and simplify the design and construction with considerable engineering application value. (authors)

  5. Seismic dynamic analysis of Heat Exchangers inside of the Auxiliary Buildings in AP1000TM NPP

    International Nuclear Information System (INIS)

    Di Fonzo, M.; Aragon, J.; Moraleda, F.; Palazuelos, M.; San Vicente, J. L.

    2011-01-01

    Seismic dynamic analysis was carried out for the Heat Exchangers (RNS-HR) located inside of the Auxiliary Building in AP 1000 T M NPP. The main function of the RNS-HX is to provide shutdown reactor cooling. These equipment's are safety-related. So the seismic analysis was done using the methodology for Seismic Category I (SCI) structures. The most important topic is that the RNS-HX shall withstand the effects of the Safe Shutdown Earthquake (SSE) and maintain the specified design functions. for the analysis, two finite element models (FEM) were built in order to investigate the structural response of the couple system of building and equipment. The response spectra method was used. The floor response spectra (FRS) at the slab-wall connection were used as input Lateral seismic restrain was necessary to added in order to achieve the natural frequency of 33 Hz. The global structural response was obtained by means of the modal combination method indicated in the Regulatory Guide 1.92.

  6. Invariant methods for an ensemble-based sensitivity analysis of a passive containment cooling system of an AP1000 nuclear power plant

    International Nuclear Information System (INIS)

    Di Maio, Francesco; Nicola, Giancarlo; Borgonovo, Emanuele; Zio, Enrico

    2016-01-01

    Sensitivity Analysis (SA) is performed to gain fundamental insights on a system behavior that is usually reproduced by a model and to identify the most relevant input variables whose variations affect the system model functional response. For the reliability analysis of passive safety systems of Nuclear Power Plants (NPPs), models are Best Estimate (BE) Thermal Hydraulic (TH) codes, that predict the system functional response in normal and accidental conditions and, in this paper, an ensemble of three alternative invariant SA methods is innovatively set up for a SA on the TH code input variables. The ensemble aggregates the input variables raking orders provided by Pearson correlation ratio, Delta method and Beta method. The capability of the ensemble is shown on a BE–TH code of the Passive Containment Cooling System (PCCS) of an Advanced Pressurized water reactor AP1000, during a Loss Of Coolant Accident (LOCA), whose output probability density function (pdf) is approximated by a Finite Mixture Model (FMM), on the basis of a limited number of simulations. - Highlights: • We perform the reliability analysis of a passive safety system of Nuclear Power Plant (NPP). • We use a Thermal Hydraulic (TH) code for predicting the NPP response to accidents. • We propose an ensemble of Invariant Methods for the sensitivity analysis of the TH code • The ensemble aggregates the rankings of Pearson correlation, Delta and Beta methods. • The approach is tested on a Passive Containment Cooling System of an AP1000 NPP.

  7. 76 FR 44377 - Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on AP1000...

    Science.gov (United States)

    2011-07-25

    ..., 2011-1 p.m. until 5 p.m. The Subcommittee will review technical updates in Revision 19 to the AP1000 Design Control Document (DCD). The Subcommittee will hear presentations by and hold discussions with the... statements can be obtained from the website cited above or by contacting the identified DFO. Moreover, in...

  8. The change of radial power factor distribution due to RCCA insertion at the first cycle core of AP1000

    Science.gov (United States)

    Susilo, J.; Suparlina, L.; Deswandri; Sunaryo, G. R.

    2018-02-01

    The using of a computer program for the PWR type core neutronic design parameters analysis has been carried out in some previous studies. These studies included a computer code validation on the neutronic parameters data values resulted from measurements and benchmarking calculation. In this study, the AP1000 first cycle core radial power peaking factor validation and analysis were performed using CITATION module of the SRAC2006 computer code. The computer code has been also validated with a good result to the criticality values of VERA benchmark core. The AP1000 core power distribution calculation has been done in two-dimensional X-Y geometry through ¼ section modeling. The purpose of this research is to determine the accuracy of the SRAC2006 code, and also the safety performance of the AP1000 core first cycle operating. The core calculations were carried out with the several conditions, those are without Rod Cluster Control Assembly (RCCA), by insertion of a single RCCA (AO, M1, M2, MA, MB, MC, MD) and multiple insertion RCCA (MA + MB, MA + MB + MC, MA + MB + MC + MD, and MA + MB + MC + MD + M1). The maximum power factor of the fuel rods value in the fuel assembly assumedapproximately 1.406. The calculation results analysis showed that the 2-dimensional CITATION module of SRAC2006 code is accurate in AP1000 power distribution calculation without RCCA and with MA+MB RCCA insertion.The power peaking factor on the first operating cycle of the AP1000 core without RCCA, as well as with single and multiple RCCA are still below in the safety limit values (less then about 1.798). So in terms of thermal power generated by the fuel assembly, then it can be considered that the AP100 core at the first operating cycle is safe.

  9. Analysis of AP1000{sup TM} reactor vessel cavity and support cooling

    Energy Technology Data Exchange (ETDEWEB)

    Craig, K.J. [Westinghouse Electric South Africa, 32 Park Avenue North, Highway Business Park, Centurion, 0157 (SOUTH AFRICA); Harkness, A.W. [Nuclear Power Plants, Westinghouse Electric Company, LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States); Kritzinger, H.P.; Hoffmann, J.E. [Pebble Bed Modular Reactor (Pty) Ltd, 1279 Mike Crawford Avenue, Centurion (South Africa)

    2010-07-01

    The paper investigates a Computational Fluid Dynamic (CFD) analysis of the air cooling of the Reactor Vessel (RV) cavity and RV supports. All the Heating, Ventilation and Air Conditioning (HVAC) flow of the RV cavity has to pass through the four RV supports supporting the four cold legs (cold inlets from the two steam generators) of the AP1000{sup TM} reactor. The RV support has a complex flow path leading to significant pressure drops to provide the necessary cooling. The insulation surrounding the RV has a specification on the amount of heat that may be transferred (lost) from the RV in order to maximize the heat transfer to the coolant driving the steam generators. This heat loss is applied as a boundary condition to the solution domain. Another heat source that is considered is that due to nuclear heating. Due to the fact that the heat source is nuclear in nature, gamma and neutron heating have to be considered for the surrounding structures. These include the carbon steel structural module that encapsulates the RV cavity, as well as the concrete poured around this module. The space in the gap between the RV insulation and the structural module steel shell is not only obstructed by the insulation supports, but also by wells or tubes within which power and intermediate ex-core detectors are located. Source-range ex-core detectors are embedded in the concrete surrounding the structural module. All these detectors have a limited operating temperature range, and together with limits on concrete temperatures for safety considerations, necessitate the need for CFD simulations to determine the range of operational temperatures seen by these components. The CFD simulations also provide an estimate of the pressure drop through the cavity between the RV insulation and structural module, as well as that through the four RV supports. Results presented include ANSYS{sup R} FLUENT{sup R} simulations describing the modelling procedure that was followed, namely to combine

  10. The preliminary analysis of establishing the cost control system of AP1000 for the Haiyang nuclear power project

    International Nuclear Information System (INIS)

    Li Jing; Li Xiaobing

    2012-01-01

    The AP1000 technology has been first applied to Nuclear Power Plant construction in China. Haiyang Project is the second plant which applies the new technology, and it is the key to the success of the project, that how to control the cost. The cost control of AP1000 is to manage and monitor all the cost of the project, including the cost of project management, design, procurement, construction, and startup/commissioning. For the Haiyang Project, the cost control system should be established to ensure that the evaluation of the procurement order should be covered in the original budget, and all potential commitments are evaluated and approved within the confinement of cost control, and reduce the risk of the first reactor and get the most profit. (authors)

  11. Fuel burn-up distribution and transuranic nuclide contents produced at the first cycle operation of AP1000

    International Nuclear Information System (INIS)

    Jati Susilo; Jupiter Sitorus Pane

    2016-01-01

    AP1000 reactor core was designed with nominal power of 1154 MWe (3415 MWth), operated within life time of 60 years and cycle length of 18 months. For the first cycle, the AP1000 core uses three kinds of UO 2 enrichment, they are 2.35 w/o, 3.40 w/o and 4.45 w/o. Absorber materials such as ZrB 2 , Pyrex and Boron solution are used to compensate the excess reactivity at the beginning of cycle. In the core, U-235 fuels are burned by fission reaction and produce energy, fission products and new neutron. Because of the U-238 neutron absorption reaction, the high level radioactive waste of heavy nuclide transuranic such as Pu, Am, Cm and Np are also generated. They have a very long half life. The purpose of this study is to evaluate the result of fuel burn-up distribution and heavy nuclide transuranic contents produced by AP1000 at the end of first cycle operation (EOFC). Calculation of ¼ part of the AP1000 core in the 2 dimensional model has been done using SRAC2006 code with the module of COREBN/HIST. The input data called the table of macroscopic cross section, is calculated using module of PIJ. The result shows that the maximum fuel assembly (FA) burn-up is 27.04 GWD/MTU, that is still lower than allowed maximum burn-up of 62 GWD/MTU. Fuel loading position at the center/middle of the core will produce bigger burn-up and transuranic nuclide than one at the edges the of the core. The use of IFBA fuel just give a small effect to lessen the fuel burn-up and transuranic nuclide production. (author)

  12. AP1000 Shield Building Dynamic Response for Different Water Levels of PCCWST Subjected to Seismic Loading considering FSI

    Directory of Open Access Journals (Sweden)

    Daogang Lu

    2015-01-01

    Full Text Available Huge water storage tank on the top of many buildings may affect the safety of the structure caused by fluid-structure interaction (FSI under the earthquake. AP1000 passive containment cooling system water storage tank (PCCWST placed at the top of shield building is a key component to ensure the safety of nuclear facilities. Under seismic loading, water will impact the wall of PCCWST, which may pose a threat to the integrity of the shield building. In the present study, an FE model of AP1000 shield building is built for the modal and transient seismic analysis considering the FSI. Six different water levels in PCCWST were discussed by comparing the modal frequency, seismic acceleration response, and von Mises stress distribution. The results show the maximum von Mises stress emerges at the joint of shield building roof and water around the air inlet. However, the maximum von Mises stress is below the yield strength of reinforced concrete. The results may provide a reference for design of the AP1000 and CAP1400 in the future.

  13. Comparative study on aerosol removal by natural processes in containment in severe accident for AP1000 reactor

    International Nuclear Information System (INIS)

    Sun, Xiaohui; Cao, Xinrong; Shi, Xingwei; Yan, Jin

    2017-01-01

    Highlights: • Characteristics of aerosol distribution in containment are obtained. • Aerosol removal by natural processes is comparative studied by two methods. • Traditional rapid assessment method is conservative and can be applied in AP1000 reactor. - Abstract: Focusing on aerosol removal by naturally occurring processes in containment in severe accident for AP1000, integral severe accident code MELCOR and rapid assessment method mentioned in NUREG/CR-6189 are utilized to study aerosol removal by natural processes, respectively. Three typical severe accidents, induced by large break loss of coolant accident (LBLOCA), small break loss of coolant accident (SBLOCA) and steam generator tube rupture (SGTR), respectively, are selected for the study. The results obtained by two methods were further compared in the following several aspects: efficiency of aerosol removal by natural processes, peak time of aerosol suspended in containment atmosphere, peak amount of aerosol suspended in containment atmosphere, time when aerosol removal efficiency by natural processes is up to 99.9%. It was further concluded that results obtained by rapid assessment with shorter calculation process are more conservative. The analysis results provide reference to assessment method selection of severe accident source term for AP1000 nuclear emergency.

  14. Analysis of the passive heat removal enhancement for AP1000 containment due to the partially wetted coverage

    Energy Technology Data Exchange (ETDEWEB)

    Li, Cheng, E-mail: 510395453@qq.com [State Nuclear Power Technology Research & Development Center, 102209 Beijing (China); Li, Le [Tsinghua University, Institute of Nuclear and New Energy Technology, 100084 Beijing (China); Li, Junming [Tsinghua University, Key Laboratory for Thermal Science and Power Engineering of Ministry of Education, Department of Thermal Engineering, Beijing 100084 (China); Zhang, Yajun [Tsinghua University, Institute of Nuclear and New Energy Technology, 100084 Beijing (China); Li, Zhihui [State Nuclear Power Technology Research & Development Center, 102209 Beijing (China)

    2017-03-15

    Highlights: • Heat removal by steam condensation, thermal conduction and evaporation is the most important scheme for AP1000 PCCS. Traditionally, studies on containment wall condensation and evaporation have been widely made, while it lacks studies on the shell two-dimension (2-D) thermal conduction. Currently, based on the known heat and mass transfer correlations and the phenomenon from water wetted coverage test, the physical model for 2-D thermal conduction is given and numerical simulation is then made. By discussions, it forms the following highlights. • The partially wetted surface can enhance the whole heat transfer process (including inner condensation, wall thermal conduction and outside cooling) and the maximum enhancement factor can be as large as 63%. There is an enhancement peak at around dry strip fraction a = 90%. When L is less than 0.03 m, its influence on heat transfer is small and the enhancement is mainly affected by dry coverage. However, for larger L, both α and L contribute much to larger enhancement. • Location at the spring line is often used for safety analysis and the dry strip fraction there for AP1000 is mainly at 10%–80%. Accordingly, further analysis is made on L (0.03 < L < 0.3) and a fitting expression is given for α = 10%–80%. It could be used to improve the corresponding software and it could also be used for containment scaling-down criteria analysis. - Abstract: AP1000 containment uses the water film evaporation, coupled with containment inner condensation, to remove the core decay heat. However, water film cannot fully cover heat transfer surface and dry-wetted strips appear. As a result, heat transfer within the containment shell is a two-dimension thermal conduction. Current work numerically studied the AP1000 heat removal enhancement due to the partially wetted coverage phenomenon. It used the evaporation and condensation boundary conditions and Fluent software to calculate the local heat fluxes and their

  15. Analysis of the passive heat removal enhancement for AP1000 containment due to the partially wetted coverage

    International Nuclear Information System (INIS)

    Li, Cheng; Li, Le; Li, Junming; Zhang, Yajun; Li, Zhihui

    2017-01-01

    Highlights: • Heat removal by steam condensation, thermal conduction and evaporation is the most important scheme for AP1000 PCCS. Traditionally, studies on containment wall condensation and evaporation have been widely made, while it lacks studies on the shell two-dimension (2-D) thermal conduction. Currently, based on the known heat and mass transfer correlations and the phenomenon from water wetted coverage test, the physical model for 2-D thermal conduction is given and numerical simulation is then made. By discussions, it forms the following highlights. • The partially wetted surface can enhance the whole heat transfer process (including inner condensation, wall thermal conduction and outside cooling) and the maximum enhancement factor can be as large as 63%. There is an enhancement peak at around dry strip fraction a = 90%. When L is less than 0.03 m, its influence on heat transfer is small and the enhancement is mainly affected by dry coverage. However, for larger L, both α and L contribute much to larger enhancement. • Location at the spring line is often used for safety analysis and the dry strip fraction there for AP1000 is mainly at 10%–80%. Accordingly, further analysis is made on L (0.03 < L < 0.3) and a fitting expression is given for α = 10%–80%. It could be used to improve the corresponding software and it could also be used for containment scaling-down criteria analysis. - Abstract: AP1000 containment uses the water film evaporation, coupled with containment inner condensation, to remove the core decay heat. However, water film cannot fully cover heat transfer surface and dry-wetted strips appear. As a result, heat transfer within the containment shell is a two-dimension thermal conduction. Current work numerically studied the AP1000 heat removal enhancement due to the partially wetted coverage phenomenon. It used the evaporation and condensation boundary conditions and Fluent software to calculate the local heat fluxes and their

  16. Feasibility to convert an advanced PWR from UO2 to a mixed (U,Th)O2 core

    International Nuclear Information System (INIS)

    Stefani, Giovanni Laranjo de; Maiorino, José Rubens; Moreira, João Manoel de Losada; Santos, Thiago Augusto dos; Rossi, Pedro Carlos Russo

    2017-01-01

    This work presents the neutronics and thermal hydraulics feasibility to convert the UO2 core of the Westinghouse AP1000 in a (U-Th)O 2 core, rather than the traditional uranium dioxide, for the purpose of reducing long-lived actinides, especially plutonium, and generates a stock pile of 233 U, which could in the future be used in advanced fuel cycles, in a more sustainable process and taking advantage of the large stock of thorium available on the planet and especially in Brazil. The reactor chosen as reference was the AP1000, which is considered to be one of the most reliable and modern reactor of the current Generation III, and its similarity to the reactors already consolidated and used in Brazil for electric power generation. The results show the feasibility and potentiality of the concept, without the necessity of changes in the core of the AP1000, and even with advantages over this. The neutron calculations were made by the SERPENT code. The results provided a maximum linear power density lower than the AP1000, favoring safety. In addition, the delayed neutron fraction and the reactivity coefficients proved to be adequate to ensure the safety of the concept. The results show that a production of about 260 Kg of 233 U per cycle is possible, with a minimum production of fissile plutonium that favors the use of the concept in U-Th cycles. (author)

  17. A neutronic feasibility study of the AP1000 design loaded with fully ceramic micro-encapsulated fuel

    International Nuclear Information System (INIS)

    Liang, C.; Ji, W.

    2013-01-01

    A neutronic feasibility study is performed to evaluate the utilization of fully ceramic microencapsulated (FCM) fuel in the AP1000 reactor design. The widely used Monte Carlo code MCNP is employed to perform the full core analysis at the beginning of cycle (BOC). Both the original AP1000 design and the modified design with the replacement of uranium dioxide fuel pellets with FCM fuel compacts are modeled and simulated for comparison. To retain the original excess reactivity, ranges of fuel particle packing fraction and fuel enrichment in the FCM fuel design are first determined. Within the determined ranges, the reactor control mechanism employed by the original design is directly used in the modified design and the utilization feasibility is evaluated. The worth of control of each type of fuel burnable absorber (discrete/integral fuel burnable absorbers and soluble boron in primary coolant) is calculated for each design and significant differences between the two designs are observed. Those differences are interpreted by the fundamental difference of the fuel form used in each design. Due to the usage of silicon carbide as the matrix material and the fuel particles fuel form in FCM fuel design, neutron slowing down capability is increased in the new design, leading to a much higher thermal spectrum than the original design. This results in different reactivity and fission power density distributions in each design. We conclude that a direct replacement of fuel pellets by the FCM fuel in the AP1000 cannot retain the original optimum reactor core performance. Necessary modifications of the core design should be done and the original control mechanism needs to be re-designed. (authors)

  18. Feasibility to convert an advanced PWR from UO{sub 2} to a mixed (U,Th)O{sub 2} core

    Energy Technology Data Exchange (ETDEWEB)

    Stefani, Giovanni Laranjo de; Maiorino, José Rubens; Moreira, João Manoel de Losada; Santos, Thiago Augusto dos, E-mail: giovanni_laranjo@yahoo.com.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); Rossi, Pedro Carlos Russo [Department of Energy, System, Territory, and Construction Engineering (DESTEC), Pisa (Italy)

    2017-07-01

    This work presents the neutronics and thermal hydraulics feasibility to convert the UO2 core of the Westinghouse AP1000 in a (U-Th)O{sub 2} core, rather than the traditional uranium dioxide, for the purpose of reducing long-lived actinides, especially plutonium, and generates a stock pile of {sup 233}U, which could in the future be used in advanced fuel cycles, in a more sustainable process and taking advantage of the large stock of thorium available on the planet and especially in Brazil. The reactor chosen as reference was the AP1000, which is considered to be one of the most reliable and modern reactor of the current Generation III, and its similarity to the reactors already consolidated and used in Brazil for electric power generation. The results show the feasibility and potentiality of the concept, without the necessity of changes in the core of the AP1000, and even with advantages over this. The neutron calculations were made by the SERPENT code. The results provided a maximum linear power density lower than the AP1000, favoring safety. In addition, the delayed neutron fraction and the reactivity coefficients proved to be adequate to ensure the safety of the concept. The results show that a production of about 260 Kg of {sup 233}U per cycle is possible, with a minimum production of fissile plutonium that favors the use of the concept in U-Th cycles. (author)

  19. TRAC analysis of an 80% pump-side, cold-leg, large-break loss-of-coolant accident for the Westinghouse AP600 advanced reactor design

    International Nuclear Information System (INIS)

    Lime, J.F.; Boyack, B.E.

    1996-01-01

    An updated TRAC 80% pump-side, cold-leg, large-break (LB) loss-of-coolant accident (LOCA) has been calculated for the Westinghouse AP600 advanced reactor design. The updated calculation incorporates major code error corrections, model corrections, and plant design changes. The break size and location were calculated by Westinghouse to be the most severe LBLOCA for the AP600 design. The LBLOCA transient was calculated to 280 s, which is the time of in-containment refueling water-storage-tank injection. All fuel rods were quenched completely by 240 s. Peak cladding temperatures (PCTs) were well below the licensing limit of 1,478 K (2,200 F) but were very near the cladding oxidation temperature of 1,200 K (1,700 F). Transient event times and PCTs for the TRAC calculation were in reasonable agreement with those calculated by Westinghouse using their WCOBRA/TRAC code. However, there were significant differences in the detailed phenomena calculated by the two codes, particularly during the blowdown and refill periods. The reasons for these differences are still being investigated

  20. FSI effects and seismic performance evaluation of water storage tank of AP1000 subjected to earthquake loading

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Chunfeng, E-mail: zhaowindy@126.com [Institute of Earthquake Engineering, Dalian University of Technology, Dalian 116024 (China); School of Civil Engineering, Hefei University of Technology, Anhui Province 230009 (China); Chen, Jianyun; Xu, Qiang [Institute of Earthquake Engineering, Dalian University of Technology, Dalian 116024 (China)

    2014-12-15

    Graphical abstract: - Highlights: • Water sloshing and oscillation of water tank under earthquake are simulated by FEM. • The influences of various water levels on seismic response are investigated. • ALE algorithm is applied to study the fluid–structure interaction effects. • The effects of different water levels in reducing seismic response are compared. • The optimal water level of water tank under seismic loading is obtained. - Abstract: The gravity water storage tank of AP1000 is designed to cool down the temperature of containment vessel by spray water when accident releases mass energy. However, the influence of fluid–structure interaction between water and water tank of AP1000 on dynamic behavior of shield building is still a hot research question. The main objective of the current study is to investigate how the fluid–structure interaction affects the dynamic behavior of water tank and whether the water sloshing and oscillation can reduce the seismic response of the shield building subjected to earthquake. For this purpose, a fluid–structure interaction algorithm of finite element technique is employed for the seismic analysis of water storage tank of AP1000. In the finite element model, 8 cases height of water, such as 10.8, 9.8, 8.8, 7.8, 6.8, 5.8, 4.8, and 3.8 m, are established and compared with the empty water tank in order to demonstrate the positive effect in mitigating the seismic response. An Arbitrary Lagrangian Eulerian (ALE) algorithm is used to simulate the fluid–structure interaction, fluid sloshing and oscillation of water tank under the El-Centro earthquake. The correlation between seismic response and parameters of water tank in terms of height of air (h{sub 1}), height of water (h{sub 2}), height ratio of water to tank (h{sub 2}/H{sub w}) and mass ratio of water to total structure (m{sub w}/m{sub t}) is also analyzed. The numerical results clearly show that the optimal h{sub 2}, h{sub 2}/H{sub w} and m{sub w}/m{sub t

  1. Management and integration of engineering and construction activities: Lessons learned from the AP1000R nuclear power plant China project

    International Nuclear Information System (INIS)

    McCullough, M. C.; Ebeling-Koning, D.; Evans, M. C.

    2012-01-01

    The lessons learned during the early phase of design engineering and construction activities for the AP1000 China Project can be applied to any project involving multiple disciplines and multiple organizations. Implementation of a first-of-a-kind design to directly support construction activities utilizing resources assigned to design development and design delivery creates challenges with prioritization of activities, successful closure of issues, and communication between site organizations and the home office. To ensure successful implementation, teams were assigned and developed to directly support construction activities including prioritization of activities, site communication and ensuring closure of site emergent issues. By developing these teams, the organization is better suited to meet the demands of the construction schedule while continuing with design evolution of a standard plant and engineering delivery for multiple projects. For a successful project, proper resource utilization and prioritization are key for overcoming obstacles and ensuring success of the engineering organization. (authors)

  2. Numerical simulation and experimental verification of microstructure evolution in large forged pipe used for AP1000 nuclear power plants

    International Nuclear Information System (INIS)

    Wang, Shenglong; Yang, Bin; Zhang, Mingxian; Wu, Huanchun; Peng, Jintao; Gao, Yang

    2016-01-01

    Highlights: • Establish systematically the database of 316LN stainless steel for Deform-3D. • Simulate the microstructure evolution during forging of AP1000 primary coolant pipe. • Carry out full-scale forging experiment for verification in engineering practice. • Get desirable grain size in simulation and experiment. • The variation trends of grain sizes in simulation and experiment are consistent. - Abstract: AP1000 primary coolant pipe is a large special-shaped forged pipe made of 316LN stainless steel. Due to the non-uniform temperature and deformation during its forging, coarse and fine grains usually coexist in the forged pipe, resulting in the heterogeneous microstructure and anisotropic performance. To investigate the microstructure evolution during the entire forging process, in the present research, the database of the 316LN stainless steel was established and a numerical simulation was performed. The results indicate that the middle body section of the forged pipe has an extremely uniform average grain size with the value smaller than 30 μm. The grain sizes in the ends of body sections were ranged from 30 μm to 60 μm. Boss sections have relatively homogeneous microstructure with the average grain size 30 μm to 44 μm. Furthermore, a full-scale hot forging was carried out for verification. Comparison of theoretical and experimental results showed good agreement and hence demonstrated the capabilities of the numerical simulation presented here. It is noteworthy that all grains in the workpiece were confirmed less than 180 μm, which meets the designer’s demands.

  3. Definition of parameters for a test section for the analysis of natural convection and coolant loss in the AP1000 nuclear reactor by similarity laws and fractional scaling analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cadiz, Luis Felipe S.; Bezerra, Mario Augusto [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear; Lima, Fernando Roberto A., E-mail: falima@cnen.gov.br [Centro Regional de Ciências Nucleares do Nordeste (CRCN-NE/CNEN-PB), Recife, PB (Brazil)

    2017-07-01

    The present work develops and analyzes the main parameters of a test section for natural convection in case of a failure of the pumping system as much as the loss of coolant in refrigeration accidents. For this realization, a combination of laws of basic similarity and an innovative scale methodology, known as Fractional Scaling Analysis (FSA), was developed. The depressurizing is analyzed when a rupture occurs in one of the primary system piping of the AP1000 nuclear reactor. This reactor is developed by Westinghouse Electric Co., which is a PWR (Pressurized Water Reactor) with an electric power equal to 1000MW. Such a reactor is provided with a passive safety system that promotes considerable improvements in the safety, reliability, protection and reduction of costs of a nuclear power plant. The FSA is based on two concepts: fractional scale and hierarchy. It is used to provide experimental data that generate quantitative evaluation criteria as well as operational parameters in thermal and hydraulic processes of nuclear power plants. The results were analyzed with the use of computational codes. (author)

  4. Knowledge-based software design for Defense-in-Depth risk monitor system and application for AP1000

    International Nuclear Information System (INIS)

    Ma Zhanguo; Yoshikawa, Hidekazu; Yang Ming; Nakagawa, Takashi

    2017-01-01

    As part of the new risk monitor system, the software for the plant Defense-in-Depth (DiD) risk monitor system was designed based on the state-transition and finite-state machine, and then the knowledge-based software was developed by object-oriented method utilizing the Unified Modeling Language (UML). Currently, there are mainly two functions in the developed plant DiD risk monitor software that are knowledge-base editor which is used to model the system in a hierarchical manner and the interaction simulator that simulates the interactions between the different actors in the model. In this paper, a model for playing its behavior is called an Actor which is modeled at the top level. The passive safety AP1000 power plant was studied and the small-break loss-of-coolant accident (SBLOCA) design basis accident transient is modeled using the plant DiD risk monitor software. Furthermore, the simulation result is shown for the interactions between the actors which are defined in the plant DiD risk monitor system as PLANT actor, OPERATOR actor, and SUPERVISOR actor. This paper shows that it is feasible to model the nuclear power plant knowledge base using the software modeling technique. The software can make the large knowledge base for the nuclear power plant with small effort. (author)

  5. Seismic dynamic analysis of Heat Exchangers inside of the Auxiliary Buildings in AP1000{sup T}M NPP

    Energy Technology Data Exchange (ETDEWEB)

    Di Fonzo, M.; Aragon, J.; Moraleda, F.; Palazuelos, M.; San vicente, J. L.

    2011-07-01

    Seismic dynamic analysis was carried out for the Heat Exchangers (RNS-HR) located inside of the Auxiliary Building in AP 1000{sup T}M NPP. The main function of the RNS-HX is to provide shutdown reactor cooling. These equipment's are safety-related. So the seismic analysis was done using the methodology for Seismic Category I (SCI) structures. The most important topic is that the RNS-HX shall withstand the effects of the Safe Shutdown Earthquake (SSE) and maintain the specified design functions. for the analysis, two finite element models (FEM) were built in order to investigate the structural response of the couple system of building and equipment. The response spectra method was used. The floor response spectra (FRS) at the slab-wall connection were used as input Lateral seismic restrain was necessary to added in order to achieve the natural frequency of 33 Hz. The global structural response was obtained by means of the modal combination method indicated in the Regulatory Guide 1.92.

  6. Design of a decoupled AP1000 reactor core control system using digital proportional–integral–derivative (PID) control based on a quasi-diagonal recurrent neural network (QDRNN)

    International Nuclear Information System (INIS)

    Wei, Xinyu; Wang, Pengfei; Zhao, Fuyu

    2016-01-01

    Highlights: • We establish a disperse dynamic model for AP1000 reactor core. • A digital PID control based on QDRNN is used to design a decoupling control system. • The decoupling performance is verified and discussed. • The decoupling control system is simulated under the load following operation. - Abstract: The control system of the AP1000 reactor core uses the mechanical shim (MSHIM) strategy, which includes a power control subsystem and an axial power distribution control subsystem. To address the strong coupling between the two subsystems, an interlock between the two subsystems is used, which can only alleviate but not eliminate the coupling. Therefore, sometimes the axial offset (AO) cannot be controlled tightly, and the flexibility of load-following operation is limited. Thus, the decoupling of the original AP1000 reactor core control system is the focus of this paper. First, a two-node disperse dynamic model is established for the AP1000 reactor core to use PID control. Then, a digital PID control system based on a quasi-diagonal recurrent neural network (QDRNN) is designed to decouple the original system. Finally, the decoupling of the control system is verified by the step signal and load-following condition. The results show that the designed control system can decouple the original system as expected and the AO can be controlled much more tightly. Moreover, the flexibility of the load following is increased.

  7. Design of a decoupled AP1000 reactor core control system using digital proportional–integral–derivative (PID) control based on a quasi-diagonal recurrent neural network (QDRNN)

    Energy Technology Data Exchange (ETDEWEB)

    Wei, Xinyu, E-mail: xyuwei@mail.xjtu.edu.cn; Wang, Pengfei, E-mail: pengfeixiaoli@yahoo.cn; Zhao, Fuyu, E-mail: fuyuzhao_xj@163.com

    2016-08-01

    Highlights: • We establish a disperse dynamic model for AP1000 reactor core. • A digital PID control based on QDRNN is used to design a decoupling control system. • The decoupling performance is verified and discussed. • The decoupling control system is simulated under the load following operation. - Abstract: The control system of the AP1000 reactor core uses the mechanical shim (MSHIM) strategy, which includes a power control subsystem and an axial power distribution control subsystem. To address the strong coupling between the two subsystems, an interlock between the two subsystems is used, which can only alleviate but not eliminate the coupling. Therefore, sometimes the axial offset (AO) cannot be controlled tightly, and the flexibility of load-following operation is limited. Thus, the decoupling of the original AP1000 reactor core control system is the focus of this paper. First, a two-node disperse dynamic model is established for the AP1000 reactor core to use PID control. Then, a digital PID control system based on a quasi-diagonal recurrent neural network (QDRNN) is designed to decouple the original system. Finally, the decoupling of the control system is verified by the step signal and load-following condition. The results show that the designed control system can decouple the original system as expected and the AO can be controlled much more tightly. Moreover, the flexibility of the load following is increased.

  8. The whiteStar development project: Westinghouse's next generation core design simulator and core monitoring software to power the nuclear renaissance

    International Nuclear Information System (INIS)

    Boyd, W. A.; Mayhue, L. T.; Penkrot, V. S.; Zhang, B.

    2009-01-01

    The WhiteStar project has undertaken the development of the next generation core analysis and monitoring system for Westinghouse Electric Company. This on-going project focuses on the development of the ANC core simulator, BEACON core monitoring system and NEXUS nuclear data generation system. This system contains many functional upgrades to the ANC core simulator and BEACON core monitoring products as well as the release of the NEXUS family of codes. The NEXUS family of codes is an automated once-through cross section generation system designed for use in both PWR and BWR applications. ANC is a multi-dimensional nodal code for all nuclear core design calculations at a given condition. ANC predicts core reactivity, assembly power, rod power, detector thimble flux, and other relevant core characteristics. BEACON is an advanced core monitoring and support system which uses existing instrumentation data in conjunction with an analytical methodology for on-line generation and evaluation of 3D core power distributions. This new system is needed to design and monitor the Westinghouse AP1000 PWR. This paper describes provides an overview of the software system, software development methodologies used as well some initial results. (authors)

  9. Development of a simulation platform for dynamic simulation and control studies of AP1000 nuclear steam supply system

    International Nuclear Information System (INIS)

    Wan, Jiashuang; Song, Hongbing; Yan, Shoujun; Sun, Jian; Zhao, Fuyu

    2015-01-01

    Highlights: • A fast-running simulation platform named NCAP was developed on a personal computer using MATLAB/Simulink. • Three types of typical operations, namely 10% step load change, 5%/min ramp load change and load follow were simulated. • NCAP predictions were compared with those obtained by CENTS for the load regulation transients. - Abstract: This paper presents the development, application and performance assessment of a fast-running NCAP (NSSS Control & Analysis Platform) in MATLAB/Simulink environment. First, a nodal core model, a lumped parameter dynamic steam generator model with moving boundary, a non-equilibrium two-regions-three-volumes pressurizer model, and the relevant pipe and plenum models were proposed based on the fundamental conservation of mass, energy and momentum. Then, these first order nonlinear models and the NSSS control systems were implemented in the Simulink by the predefined library blocks. Based on the developed NCAP, three types of typical operational transients, namely the 10% step load change, the 5%/min ramp load change and the daily load follow were simulated to study the dynamic behavior and control characteristics of the AP1000 NSSS. It has been demonstrated that the dynamic responses of the selected key parameters agree well with the general physical rules. In addition, the comparison of load regulation simulation results obtained by NCAP and CENTS shows a good agreement in terms of the changing trends. With the adoption of modular programming techniques, the NCAP facilitates easy modification and runs quickly, which easily allows the control system designer to test and compare various ideas efficiently

  10. Westinghouse Small Modular Reactor nuclear steam supply system design

    Energy Technology Data Exchange (ETDEWEB)

    Memmott, M. J.; Harkness, A. W.; Van Wyk, J. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam

  11. Optimization of plutonium and minor actinide transmutation in an AP1000 fuel assembly via a genetic search algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Washington, J., E-mail: jwashing@gmail.com; King, J., E-mail: kingjc@mines.edu

    2017-01-15

    Highlights: • We model a modified AP1000 fuel assembly in SCALE6.1. • We couple the NEWT module of SCALE to the MOGA module of DAKOTA. • Transmutation is optimized based on choice of coating and fuel. • Greatest transmutation achieved with PuZrO{sub 2}MgO fuel pins coated with Lu{sub 2}O{sub 3}. - Abstract: The average nuclear power plant produces twenty metric tons of used nuclear fuel per year, which contains approximately 95 wt% uranium, 1 wt% plutonium, and 4 wt% fission products and transuranic elements. Fast reactors are the preferred option for the transmutation of plutonium and minor actinides; however, an optimistic deployment time of at least 20 years indicates a need for a near-term solution. Previous simulation work demonstrated the potential to transmute transuranic elements in a modified light water reactor fuel pin. This study optimizes a quarter-assembly containing target fuels coated with spectral shift absorbers for the transmutation of plutonium and minor actinides in light water reactors. The spectral shift absorber coating on the target fuel pin tunes the neutron energy spectrum experienced by the target fuel. A coupled model developed using the NEWT module from SCALE 6.1 and a genetic algorithm module from the DAKOTA optimization toolbox provided performance data for the burnup of the target fuel pins in the present study. The optimization with the coupled NEWT/DAKOTA model proceeded in three stages. The first stage optimized a single-target fuel pin per quarter-assembly adjacent to the central instrumentation channel. The second stage evaluated a variety of quarter-assemblies with multiple target fuel pins from the first stage and the third stage re-optimized the pins in the optimal second stage quarter-assembly. An 8 wt% PuZrO{sub 2}MgO inert matrix fuel pin with a 1.44 mm radius and a 0.06 mm Lu{sub 2}O{sub 3} coating in a five target fuel pin per quarter-assembly configuration represents the optimal combination for the

  12. Preliminary LOCA analysis of the westinghouse small modular reactor using the WCOBRA/TRAC-TF2 thermal-hydraulics code

    Energy Technology Data Exchange (ETDEWEB)

    Liao, J.; Kucukboyaci, V. N.; Nguyen, L.; Frepoli, C. [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) with all primary components, including the steam generator and the pressurizer located inside the reactor vessel. The reactor core is based on a partial-height 17x17 fuel assembly design used in the AP1000{sup R} reactor core. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. A preliminary loss of coolant accident (LOCA) analysis of the Westinghouse SMR has been performed using the WCOBRA/TRAC-TF2 code, simulating a transient caused by a double ended guillotine (DEG) break in the direct vessel injection (DVI) line. WCOBRA/TRAC-TF2 is a new generation Westinghouse LOCA thermal-hydraulics code evolving from the US NRC licensed WCOBRA/TRAC code. It is designed to simulate PWR LOCA events from the smallest break size to the largest break size (DEG cold leg). A significant number of fluid dynamics models and heat transfer models were developed or improved in WCOBRA/TRAC-TF2. A large number of separate effects and integral effects tests were performed for a rigorous code assessment and validation. WCOBRA/TRAC-TF2 was introduced into the Westinghouse SMR design phase to assist a quick and robust passive cooling system design and to identify thermal-hydraulic phenomena for the development of the SMR Phenomena Identification Ranking Table (PIRT). The LOCA analysis of the Westinghouse SMR demonstrates that the DEG DVI break LOCA is mitigated by the injection and venting from the Westinghouse SMR passive safety systems without core heat up, achieving long term core cooling. (authors)

  13. The power of simplification: Operator interface with the AP1000{sup R} during design-basis and beyond design-basis events

    Energy Technology Data Exchange (ETDEWEB)

    Williams, M. G.; Mouser, M. R.; Simon, J. B. [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    The AP1000{sup R} plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and cost. The passive safety features are designed to function without safety-grade support systems such as component cooling water, service water, compressed air or HVAC. The AP1000 passive safety features achieve and maintain safe shutdown in case of a design-basis accident for 72 hours without need for operator action, meeting the expectations provided in the European Utility Requirements and the Utility Requirement Document for passive plants. Limited operator actions may be required to maintain safe conditions in the spent fuel pool (SFP) via passive means. This safety approach therefore minimizes the reliance on operator action for accident mitigation, and this paper examines the operator interaction with the Human-System Interface (HSI) as the severity of an accident increases from an anticipated transient to a design basis accident and finally, to a beyond-design-basis event. The AP1000 Control Room design provides an extremely effective environment for addressing the first 72 hours of design-basis events and transients, providing ease of information dissemination and minimal reliance upon operator actions. Symptom-based procedures including Emergency Operating Procedures (EOPs), Abnormal Operating Procedures (AOPs) and Alarm Response Procedures (ARPs) are used to mitigate design basis transients and accidents. Use of the Computerized Procedure System (CPS) aids the operators during mitigation of the event. The CPS provides cues and direction to the operators as the event progresses. If the event becomes progressively worse or lasts longer than 72 hours, and depending upon the nature of failures that may have occurred, minimal operator actions may be required outside of the control room in areas that have been designed to be accessible using components that have been

  14. The power of simplification: Operator interface with the AP1000R during design-basis and beyond design-basis events

    International Nuclear Information System (INIS)

    Williams, M. G.; Mouser, M. R.; Simon, J. B.

    2012-01-01

    The AP1000 R plant is an 1100-MWe pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and cost. The passive safety features are designed to function without safety-grade support systems such as component cooling water, service water, compressed air or HVAC. The AP1000 passive safety features achieve and maintain safe shutdown in case of a design-basis accident for 72 hours without need for operator action, meeting the expectations provided in the European Utility Requirements and the Utility Requirement Document for passive plants. Limited operator actions may be required to maintain safe conditions in the spent fuel pool (SFP) via passive means. This safety approach therefore minimizes the reliance on operator action for accident mitigation, and this paper examines the operator interaction with the Human-System Interface (HSI) as the severity of an accident increases from an anticipated transient to a design basis accident and finally, to a beyond-design-basis event. The AP1000 Control Room design provides an extremely effective environment for addressing the first 72 hours of design-basis events and transients, providing ease of information dissemination and minimal reliance upon operator actions. Symptom-based procedures including Emergency Operating Procedures (EOPs), Abnormal Operating Procedures (AOPs) and Alarm Response Procedures (ARPs) are used to mitigate design basis transients and accidents. Use of the Computerized Procedure System (CPS) aids the operators during mitigation of the event. The CPS provides cues and direction to the operators as the event progresses. If the event becomes progressively worse or lasts longer than 72 hours, and depending upon the nature of failures that may have occurred, minimal operator actions may be required outside of the control room in areas that have been designed to be accessible using components that have been designed

  15. Post-remedial-action radiological survey of the Westinghouse Advanced Reactors Division Plutonium Fuel Laboratories, Cheswick, Pennsylvania, October 1-8, 1981

    International Nuclear Information System (INIS)

    Flynn, K.F.; Justus, A.L.; Sholeen, C.M.; Smith, W.H.; Wynveen, R.A.

    1984-01-01

    The post-remedial-action radiological assessment conducted by the ANL Radiological Survey Group in October 1981, following decommissioning and decontamination efforts by Westinghouse personnel, indicated that except for the Advanced Fuels Laboratory exhaust ductwork and north wall, the interior surfaces of the Plutonium Laboratory and associated areas within Building 7 and the Advanced Fuels Laboratory within Building 8 were below both the ANSI Draft Standard N13.12 and NRC Guideline criteria for acceptable surface contamination levels. Hence, with the exceptions noted above, the interior surfaces of those areas within Buildings 7 and 8 that were included in the assessment are suitable for unrestricted use. Air samples collected at the involved areas within Buildings 7 and 8 indicated that the radon, thoron, and progeny concentrations within the air were well below the limits prescribed by the US Surgeon General, the Environmental Protection Agency, and the Department of Energy. The Building 7 drain lines are contaminated with uranium, plutonium, and americium. Radiochemical analysis of water and dirt/sludge samples collected from accessible Low-Bay, High-Bay, Shower Room, and Sodium laboratory drains revealed uranium, plutonium, and americium contaminants. The Building 7 drain lines hence are unsuitable for release for unrestricted use in their present condition. Low levels of enriched uranium, plutonium, and americium were detected in an environmental soil coring near Building 8, indicating release or spillage due to Advanced Reactors Division activities or Nuclear Fuel Division activities undr NRC licensure. 60 Co contamination was detected within the Building 7 Shower Room and in soil corings from the environs of Building 7. All other radionuclide concentrations measured in soil corings and the storm sewer outfall sample collected from the environs about Buildings 7 and 8 were within the range of normally expected background concentrations

  16. MHI - Westinghouse joint FBR tank plant design

    International Nuclear Information System (INIS)

    Arnold, W.H.; Vijuk, R.M.; Aoki, I.; Messhil, T.

    1988-01-01

    Mitsubishi Heavy Industries and Westinghouse Advanced Energy Systems Division have combined their experience and capabilities to design a tank type fast breeder reactor plant. This tank type reactor has been refined and improved during the last three years to better compete in cost, safety, and operation with alternative power plants. This Mitsubishi/Westinghouse joint design offers economic advantages due to the use of steel structures, modular construction, nitrogen cells for the intermediate loops, reactor cavity air cooling and the use of the guard vessel as the containment vessel. Inherent characteristics in the reactor design provide protection to the public and the plant investment

  17. Westinghouse European trainee program

    International Nuclear Information System (INIS)

    Jimenez, G.

    2010-01-01

    Westinghouse Electric Company is proud of giving its employees the possibility to work and act globally. The company's European Trainee Program provides an opportunity to work within different fields of business within Westinghouse, participating in a wide range of projects and experiencing and learning from the different cultures of the company. In 2006 the first Trainee Program started with seven Swedish Trainees. During these eighteen months they worked 12 months in Sweden and then went off to six-month-assignments in France and in the US. In April 2008, the first European Trainee Program was launched with ten Trainees from four different countries: five from Sweden, two from Germany, two from Spain and one from Belgium. As with the previous program, its length was eighteen months. During the first year, the European Trainees had the opportunity to work in various areas within their country of hire, as well as to visit different Westinghouse headquarters in Europe and the US to learn more about the global business. Their kick-off session took place in Vaesteraas, Sweden in April 2008. During four days, the Trainees participated in group dynamic exercises as well as presentations of the business of Westinghouse abroad and in Sweden. Two of the most interesting parts of this session were the visits to the Fuel Factory and to the Field Services mock-ups. The second session took place in June 2008 in Monroeville, Pennsylvania (USA), where Westinghouse had its main headquarters, nowadays located in Cranberry, PA. During two weeks, the trainees got to know even more about Westinghouse through visits, lectures and forums for open discussions. The visits comprised for example the tubing factory at Blairsville, the Field Services main headquarters in Madison and the George Westinghouse Research and Technology Park near Pittsburgh. The meetings included presentations of each Westinghouse business unit, detailed information about future projects and round table discussions

  18. Analysis of neutronic parameters of AP1000 core for 18 month and 16/20 month cycle schemes using CASMO4E and SIMULATE-3 codes

    International Nuclear Information System (INIS)

    Nawaz Amjad; Yoshikawa, Hidekazu; Ming Yang

    2015-01-01

    AP1000 reactor is designed for 18 month of operating cycle. The core can also be used for 16/20 months of operating cycle. This study is performed to analyze and compare the neutronic parameters of typical AP1000 reactor core for 18 month and 16/20 month alternate cycle lengths. CASMO4E and SIMULATE-3 code package is used for the analysis of initial and equilibrium cores. The key reactor physics safety parameters were analyzed including power peaking factors, core radial and axial power distribution and core reactivity feedback coefficients. Moreover, the analysis of fuel depletion, fission product buildup and burnable poison behaviour with burnup is also analyzed. Full 2-D fuel assembly model in CASMO4E and full 3-D core model in SIMULATE-3 is employed to examine core performance and safety parameters. In order to evaluate the equilibrium core neutronic parameters, the equilibrium core model is attained by performing burnup analysis from initial to equilibrium cycle, where optimized transition core design is obtained so that the power peaking factors remain within designed limits. The MTC for higher concentration of critical boron concentrations is slightly positive at lower moderator temperatures. However, it remains negative at operating temperature ranges. The radial core relative power distribution indicates that low leakage capability of initial and equilibrium cores is reduced at EOC. (author)

  19. Flow Accelerated Erosion-Corrosion (FAC) considerations for secondary side piping in the AP1000{sup R} nuclear power plant design

    Energy Technology Data Exchange (ETDEWEB)

    Vanderhoff, J. F.; Rao, G. V. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States); Stein, A. [Shaw Power Nuclear, 1000 Technology Center Drive, Stoughton, MA 02072 (United States)

    2012-07-01

    The issue of Flow Accelerated Erosion-Corrosion (FAC) in power plant piping is a known phenomenon that has resulted in material replacements and plant accidents in operating power plants. Therefore, it is important for FAC resistance to be considered in the design of new nuclear power plants. This paper describes the design considerations related to FAC that were used to develop a safe and robust AP1000{sup R} plant secondary side piping design. The primary FAC influencing factors include: - Fluid Temperature - Pipe Geometry/layout - Fluid Chemistry - Fluid Velocity - Pipe Material Composition - Moisture Content (in steam lines) Due to the unknowns related to the relative impact of the influencing factors and the complexities of the interactions between these factors, it is difficult to accurately predict the expected wear rate in a given piping segment in a new plant. This paper provides: - a description of FAC and the factors that influence the FAC degradation rate, - an assessment of the level of FAC resistance of AP1000{sup R} secondary side system piping, - an explanation of options to increase FAC resistance and associated benefits/cost, - discussion of development of a tool for predicting FAC degradation rate in new nuclear power plants. (authors)

  20. Westinghouse introduces new fuel for PWRs and BWRs

    Energy Technology Data Exchange (ETDEWEB)

    Orr, W L; McClintock, D C

    1985-09-01

    In response to utility demands for improved fuel performance, reduced fuel cycle costs, and enhanced operating margins, Westinghouse recently introduced advanced fuel assembly designs for both types of LWR - Vantage 5 for PWRs, and Quad+ for BWRs.

  1. Quality surveillance for steel forgings of SA508 Gr.3 used on the main NI equipment of AP1000 nuclear island

    International Nuclear Information System (INIS)

    Liu Lizhao

    2011-01-01

    Being a type of steel with ideal weldability, outstanding ability of anti-neutron irradiation embitterment and good property of fracture toughness and impact toughness, the steel of ASME SA508-3 was used widely for the nuclear island equipment of PWR Nuclear Power Plant. For the 3rd generation nuclear power plant AP1000, all large forgings and some critical components of the SG, RV and PRZ adopt this steel. Through analysis on the critical technical points during manufacturing of the SA508-3 forgings, this article try to identify the key points should be paid attention during the quality surveillance for this type of forgings, and to put forward the supervision method and focus during quality surveillance activities. (author)

  2. Management and integration of engineering and construction activities: Lessons learned from the AP1000{sup R} nuclear power plant China project

    Energy Technology Data Exchange (ETDEWEB)

    McCullough, M. C.; Ebeling-Koning, D.; Evans, M. C. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    The lessons learned during the early phase of design engineering and construction activities for the AP1000 China Project can be applied to any project involving multiple disciplines and multiple organizations. Implementation of a first-of-a-kind design to directly support construction activities utilizing resources assigned to design development and design delivery creates challenges with prioritization of activities, successful closure of issues, and communication between site organizations and the home office. To ensure successful implementation, teams were assigned and developed to directly support construction activities including prioritization of activities, site communication and ensuring closure of site emergent issues. By developing these teams, the organization is better suited to meet the demands of the construction schedule while continuing with design evolution of a standard plant and engineering delivery for multiple projects. For a successful project, proper resource utilization and prioritization are key for overcoming obstacles and ensuring success of the engineering organization. (authors)

  3. Westinghouse radiological containment guide

    International Nuclear Information System (INIS)

    Aitken, S.B.; Brown, R.L.; Cantrell, J.R.; Wilcox, D.P.

    1994-03-01

    This document provides uniform guidance for Westinghouse contractors on the implementation of radiological containments. This document reflects standard industry practices and is provided as a guide. The guidance presented herein is consistent with the requirements of the DOE Radiological Control Manual (DOE N 5480.6). This guidance should further serve to enable and encourage the use of containments for contamination control and to accomplish the following: Minimize personnel contamination; Prevent the spread of contamination; Minimize the required use of protective clothing and personal protective equipment; Minimize the generation of waste

  4. Westinghouse radiological containment guide

    Energy Technology Data Exchange (ETDEWEB)

    Aitken, S.B. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Brown, R.L. [Westinghouse Hanford Co., Richland, WA (United States); Cantrell, J.R. [Westinghouse Savannah River Co., Aiken, SC (United States); Wilcox, D.P. [West Valley Nuclear Services Co., Inc., West Valley, NY (United States)

    1994-03-01

    This document provides uniform guidance for Westinghouse contractors on the implementation of radiological containments. This document reflects standard industry practices and is provided as a guide. The guidance presented herein is consistent with the requirements of the DOE Radiological Control Manual (DOE N 5480.6). This guidance should further serve to enable and encourage the use of containments for contamination control and to accomplish the following: Minimize personnel contamination; Prevent the spread of contamination; Minimize the required use of protective clothing and personal protective equipment; Minimize the generation of waste.

  5. Feasibility to convert an advanced PWR from UO2 to a mixed U/ThO2 core – Part I: Parametric studies

    International Nuclear Information System (INIS)

    Maiorino, Jose R.; Stefani, Giovanni Laranjo; Moreira, João M.L.; Rossi, Pedro C.R.; Santos, Thiago A.

    2017-01-01

    Highlights: • Neutronics calculation using SERPENT code. • Conversion of an advanced PWR from a UO 2 to (U-Th)O 2 core. • AP 1000-advanced PWR. • Parametric studies to define a converted core. • Demonstration of the feasibility to convert the AP 1000 by using mixed uranium thorium oxide fuel with advantages. - Abstract: This work presents the neutronics and thermal hydraulics feasibility to convert the UO 2 core of the Westinghouse AP1000 in a (U-Th)O 2 core by performing a parametric study varying the type of geometry of the pins in fuel elements, using the heterogeneous seed blanket concept and the homogeneous concept. In the parametric study, all geometry and materials for the burnable poison were kept the same as the AP 1000, and the only variable was the fuel pin material, in which we use several mass proportion of uranium and thorium but keeping the enrichment in 235 U, as LEU (20 w/o). The neutronics calculations were made by SERPENT code, and to validate the thermal limits we used a homemade code. The optimization criteria were to maximize the 233 U, and conversion factor, and minimize the plutonium production. The results obtained showed that the homogeneous concept with three different mass proportion zones, the first containing (32% UO 2 -68%ThO 2 ); the second with (24% UO 2 -76% ThO 2 ), and the third with (20% UO 2 -80% ThO 2 ), using 235 U LEU (20 w/o), and corresponding with the 3 enrichment zones of the AP 1000 (4.45 w/o; 3.40 w/o; 2.35 w/o), satisfies the optimization criteria as well as attending all thermal constrain. The concept showed advantages compared with the original UO 2 core, such a lower power density, and keeping the same 18 months of cycle a reduction of B-10 concentration at the soluble poison as well as eliminating in the integral boron poison coated (IFBA).

  6. Master of engineering program for Westinghouse Electric Corporation

    International Nuclear Information System (INIS)

    Klevans, E.H.; Diethorn, W.S.

    1991-01-01

    In August of 1985, Westinghouse Corporation, via a grant to the nuclear engineering department at Pennsylvania State University, provided its professional employees the opportunity to earn a master of engineering (M. Eng.) degree in nuclear engineering in a program of evening study in the Pittsburgh area. Faculty members from the nuclear engineering department, which is 135 miles from Westinghouse, and adjunct faculty from the professional ranks of Westinghouse provided the instruction at the Westinghouse training center facility in Monroeville, Pennsylvania, A 3-yr 30-credit program was originally planned, but this was extended to a fourth year to accommodate the actual student progress toward the degree. A fifth year was added for students to complete their engineering paper. There have been benefits to both Westinghouse and Penn State from this program. Advanced education for its employees has met a Westinghouse need. For Penn State, there has been an increase in interaction with Westinghouse personnel, and this has now led to cooperative research programs with them

  7. A Calculation Method for the Sloshing Impact Pressure Imposed on the Roof of a Passive Water Storage Tank of AP1000

    Directory of Open Access Journals (Sweden)

    Daogang Lu

    2016-01-01

    Full Text Available There is a large water storage tank installed at the top of containment of AP1000, which can supply the passive cooling. In the extreme condition, sloshing of the free surface in the tank may impact on the roof under long-period earthquake. For the safety assessment of structure, it is necessary to calculate the impact pressure caused by water sloshing. Since the behavior of sloshing impacted on the roof is involved into a strong nonlinear phenomenon, it is a little difficult to calculate such pressure by theoretical or numerical method currently. But it is applicable to calculate the height of sloshing in a tank without roof. In the present paper, a simplified method was proposed to calculate the impact pressure using the sloshing wave height, in which we first marked the position of the height of roof, then produced sloshing in the tank without roof and recorded the maximum wave height, and finally regarded approximately the difference between maximum wave height and roof height as the impact pressure head. We also designed an experiment to verify this method. The experimental result showed that this method overpredicted the impact pressure with a certain error of no more than 35%. By the experiment, we conclude that this method is conservative and applicable for the engineering design.

  8. Status of Westinghouse coal-fueled combustion turbine programs

    International Nuclear Information System (INIS)

    Scalzo, A.J.; Amos, D.J.; Bannister, R.L.; Garland, R.V.

    1992-01-01

    Developing clean, efficient, cost effective coal utilization technologies for future power generation is an essential part of our National Energy Strategy. Westinghouse is actively developing power plants utilizing advanced gasification, atmospheric fluidized beds (AFB), pressurized fluidized beds (PFB), and direct firing technology through programs sponsored by the U.S. Dept. of Energy (DOE). The DOE Office of Fossil Energy is sponsoring the Direct Coal-Fired Turbine program. This paper presents the status of current and potential Westinghouse Power Generation Business Unit advanced coal-fueled power generation programs as well as commercial plans

  9. Westinghouse Electric. Know-how and top technology from Germany support non-polluting, safe, cost-effective power supply worldwide

    International Nuclear Information System (INIS)

    2011-01-01

    Westinghouse Electric Company LLC is one the world's leading firms in the commercial nuclear power field with a staff of approx. 15,000, of whom approx. 5,000 work in Europe. As part of the Toshiba Group, Westinghouse supports power utilities in the Americas, Asia, and EMEA (Europe, Middle East, Africa) regions with a broad range of products and services in nuclear power plants, nuclear fuel, nuclear services, and nuclear automation. The German-based company, Westinghouse Electric Germany GmbH, has more than 500 persons at the locations of Mannheim; Hamburg; Baden, Switzerland; and Metz, France. For more than 40 years, it has been successfully operating in field services, plant engineering, waste management, and nuclear automation. The Mannheim head office works the nuclear markets in Germany, Switzerland, the Czech Republic, Slovakia, and Hungary. Under global resource utilization and products schemes, staff from Germany is employed also in projects all over the world. Present construction of a large number of new plants of the AP1000 registered reactor line in China and USA as well as planning and licensing steps for the construction of new nuclear power plants in Europe constitute a major contribution by Westinghouse to the worldwide renaissance of nuclear power. As a partner of utilities, Westinghouse also upgrades existing plants by backfitting and modernizing components and systems, management of aging, safety analyses, non-destructive testing, replacement of safety and operations I and C etc. for plant life extension and safe, economically viable continued operation. (orig.)

  10. MHI-Westinghouse joint FBR tank plant design

    International Nuclear Information System (INIS)

    Arnold, W.H.; Vijuk, R.M.; Aoki, I.; Meshii, T.

    1987-01-01

    Mitsubishi Heavy Industries and Westinghouse Advanced Energy Systems Division have combined their experience and capabilities to design a tank type fast breeder reactor plant. This tank type reactor has been refined and improved during the last three years to better compete in cost, satety, and operation with alternative power plants. This Mitsubishi/Westinghouse joint design offers economic advantages due to the use of steel structures, modular construction, nitrogen cells for the intermediate loops, reactor cavity air cooling and the use of the guard vessel as the containment vessel. Inherent characteristics in the reactor design provide protection to the public and the plant investment. (author)

  11. Innovation and future in Westinghouse

    International Nuclear Information System (INIS)

    Congedo, T.; Dulloo, A.; Goosen, J.; Llovet, R.

    2007-01-01

    For the past six years, Westinghouse has used a Road Map process to direct technology development in a way that integrates the efforts of our businesses to addresses the needs of our customers and respond to significant drivers in the evolving business environment. As the nuclear industry experiences a resurgence, it is ever more necessary that we increase our planning horizon to 10-15 years in the future so as to meet the expectations of our customers. In the Future Point process, driven by the methods of Design for Six Sigma (DFSS), Westinghouse considers multiple possible future scenarios to plan long term evolutionary and revolutionary development that can reliably create the major products and services of the future market. the products and services of the future stretch the imagination from what we provide today. However, the journey to these stretch targets prompts key development milestones that will help deliver ideas useful for nearer term products. (Author) 1 refs

  12. Manufacturing development of the Westinghouse Nb3Sn coil for the Large Coil Test Program

    International Nuclear Information System (INIS)

    Young, J.L.; Vota, T.L.; Singh, S.K.

    1983-01-01

    The Westinghouse Nb 3 Sn Magnet for the Oak Ridge National Laboratory Large Coil Program (LCP) is currently well into the manufacturing phase. This paper identifies the manufacturing processes and development tasks for his unique, advanced coil

  13. Westinghouse support for Spanish nuclear industry

    International Nuclear Information System (INIS)

    Rebollo, R.

    1999-01-01

    One of the major commitments Westinghouse has with the nuclear industry is to provide to the utilities the support necessary to have their nuclear units operating at optimum levels of availability and safety. This article outlines the organization the Energy Systems Business Unit of Westinghouse has in place to fulfill this commitment and describes the evolution of the support Westinghouse is providing to the operation o f the Spanish Nuclear Power plants. (Author)

  14. Energy Conversion Alternatives Study (ECAS), Westinghouse phase 1. Volume 11: Advanced steam systems. [energy conversion efficiency for electric power plants using steam

    Science.gov (United States)

    Wolfe, R. W.

    1976-01-01

    A parametric analysis was made of three types of advanced steam power plants that use coal in order to have a comparison of the cost of electricity produced by them a wide range of primary performance variables. Increasing the temperature and pressure of the steam above current industry levels resulted in increased energy costs because the cost of capital increased more than the fuel cost decreased. While the three plant types produced comparable energy cost levels, the pressurized fluidized bed boiler plant produced the lowest energy cost by the small margin of 0.69 mills/MJ (2.5 mills/kWh). It is recommended that this plant be designed in greater detail to determine its cost and performance more accurately than was possible in a broad parametric study and to ascertain problem areas which will require development effort. Also considered are pollution control measures such as scrubbers and separates for particulate emissions from stack gases.

  15. Standard Technical Specifications, Westinghouse plants

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for Westinghouse Plants and documents the positions of the Nuclear Regulatory Commission based on the Westinghouse Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. This document, Volume 3, contains the Bases for Sections 3.4--3.9 of the improved STS

  16. Standard Technical Specifications, Westinghouse Plants

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for Westinghouse Plants and documents the positions of the Nuclear Regulatory Commission based on the Westinghouse Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the unproved STS. Volume 3 contains the Bases for Sections 3.4--3.9 of the improved STS which contain information on safety limits, reactivity control systems, power distribution limits, and instrumentation

  17. Standard Technical Specifications, Westinghouse plants

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for Westinghouse Plants and documents the positions of the Nuclear Regulatory Commission based on the Westinghouse Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. This document, Volume 1, contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.9 of the improved STS

  18. Reactor physics methods development at Westinghouse

    International Nuclear Information System (INIS)

    Mueller, E.; Mayhue, L.; Zhang, B.

    2007-01-01

    The current state of reactor physics methods development at Westinghouse is discussed. The focus is on the methods that have been or are under development within the NEXUS project which was launched a few years ago. The aim of this project is to merge and modernize the methods employed in the PWR and BWR steady-state reactor physics codes of Westinghouse. (author)

  19. APWR - Mitsubishi, Japan/Westinghouse, USA

    International Nuclear Information System (INIS)

    Aeba, Y.; Weiss, E.H.

    1999-01-01

    Nuclear power generated by light water reactors accounts for approximately 1/3 of Japan's power supply. Development of the Advanced Pressurized Water Reactor (APWR) was initiated by five PWR electric power companies (Hokkaido, Kansai, Shikoku, Kyushu and Japan Atomic Power), Mitsubishi Heavy Industries, and Westinghouse, with a view to providing a nuclear power source to meet future energy demand in Japan. The APWR was developed based on the results of the Improvement and Standardization Program, promoted by the Ministry of International Trade and Industry, with reconsideration of the needs of age, such as construction cost reduction, enhanced safety and increased reliability. One of the important concepts of the APWR is its large power rating that decreases the construction cost per unit of electric generation capacity. Though the electric output was lower at the early stage of basic design than it is now, uprating to approximately 1530 MW is achieved based on the results of design progress and high efficiency improvements to the steam turbine and reactor coolant pumps. Furthermore, the APWR remarkably enhances reliability, safety operability and maintainability by introducing new technologies that include a radial reflector and advanced accumulators. The first APWR is planned to be built at Tsuruga No. 3 and No. 4 by the Japan Atomic Power Company and will be the largest commercial operation plant in the early 21st century. (author)

  20. Westinghouse Small Modular Reactor passive safety system response to postulated events

    International Nuclear Information System (INIS)

    Smith, M. C.; Wright, R. F.

    2012-01-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor. This paper is part of a series of four describing the design and safety features of the Westinghouse SMR. This paper focuses in particular upon the passive safety features and the safety system response of the Westinghouse SMR. The Westinghouse SMR design incorporates many features to minimize the effects of, and in some cases eliminates the possibility of postulated accidents. The small size of the reactor and the low power density limits the potential consequences of an accident relative to a large plant. The integral design eliminates large loop piping, which significantly reduces the flow area of postulated loss of coolant accidents (LOCAs). The Westinghouse SMR containment is a high-pressure, compact design that normally operates at a partial vacuum. This facilitates heat removal from the containment during LOCA events. The containment is submerged in water which also aides the heat removal and provides an additional radionuclide filter. The Westinghouse SMR safety system design is passive, is based largely on the passive safety systems used in the AP1000 R reactor, and provides mitigation of all design basis accidents without the need for AC electrical power for a period of seven days. Frequent faults, such as reactivity insertion events and loss of power events, are protected by first shutting down the nuclear reaction by inserting control rods, then providing cold, borated water through a passive, buoyancy-driven flow. Decay heat removal is provided using a layered approach that includes the passive removal of heat by the steam drum and independent passive heat removal system that transfers heat from the primary system to the environment. Less frequent faults such as loss of coolant accidents are mitigated by passive injection of a large quantity of water that is readily available inside containment. An automatic depressurization system is used to

  1. Requirements management at Westinghouse Electric Company

    International Nuclear Information System (INIS)

    Gustavsson, Henrik

    2014-01-01

    Field studies and surveys made in various industry branches support the Westinghouse opinion that qualitative systems engineering and requirements management have a high value in the development of complex systems and products. Two key issues causing overspending and schedule delays in projects are underestimation of complexity and misunderstandings between the different sub-project teams. These issues often arise when a project jumps too early into detail design. Good requirements management practice before detail design helps the project teams avoid such issues. Westinghouse therefore puts great effort into requirements management. The requirements management methodology at Westinghouse rests primarily on four key cornerstones: 1 - Iterative team work when developing requirements specifications, 2 - Id number tags on requirements, 3 - Robust change routine, and 4 - Requirements Traceability Matrix. (authors)

  2. Toshiba-Westinghouse, the new electronuclear giant

    International Nuclear Information System (INIS)

    Guezel, J.Ch.

    2006-01-01

    Toshiba, so far a minor actor of the world nuclear industry, won in summer 2005 in front of General Electric and Mitsubishi Heavy Industries, the takeover bid launched by the public British organization BNFL which controls Westinghouse. In case of success of this operation, Toshiba will own a quarter of the world nuclear capacities and will become the first competitor of Areva. The main objective of Toshiba is to win market shares abroad thanks to the prospects offered by Westinghouse's technologies in particular in China which is one of the most targeted market today. Short paper. (J.S.)

  3. Quality assurance plan, Westinghouse Water Reactor Divisions

    Energy Technology Data Exchange (ETDEWEB)

    1976-03-01

    The Quality Assurance Program used by Westinghouse Nuclear Energy Systems Water Reactor Divisions is described. The purpose of the program is to assure that the design, materials, and workmanship on Nuclear Steam Supply System (NSSS) equipment meet applicable safety requirements, fulfill the requirements of the contracts with the applicants, and satisfy the applicable codes, standards, and regulatory requirements.

  4. Westinghouse-DOE integration: Meeting the challenge

    International Nuclear Information System (INIS)

    Price, S.V.

    1992-01-01

    The Westinghouse Electric Corporation (WEC) is in a unique position to affect national environmental management policy approaching the 21st Century. Westinghouse companies are management and operating contractors (MOC,s) at several environmentally pivotal government-owned, contractor operated (GOCO) facilities within the Department of Energy's (DOE's) nuclear defense complex. One way the WEC brings its companies together is by activating teams to solve specific DOE site problems. For example, one challenging issue at DOE facilities involves the environmentally responsible, final disposal of transuranic and high-level nuclear wastes (TRUs and HLWS). To address these disposal issues, the DOE supports two Westinghouse-based task forces: The TRU Waste Acceptance Criteria Certification Committee (WACCC) and the HLW Vitrification Committee. The WACCC is developing methods to characterize an estimated 176,287 cubic meters of retrievably stored TRUs generated at DOE production sites. Once characterized, TRUs could be safely deposited in the WIPP repository. The Westinghouse HLW Vitrification Committee is dedicated to assess appropriate methods to process an estimated 380,702 cubic meters of HLWs currently stored in underground storage tanks (USTs). As planned, this processing will involve segregating, and appropriately treating, low level waste (LLW) and HLW tank constituents for eventual disposal. The first unit designed to process these nuclear wastes is the SRS Defense Waste Processing Facility (DWPF). Initiated in 1973, the DWPF project is scheduled to begin operations in 1991 or 1992. Westinghouse companies are also working together to achieve appropriate environmental site restoration at DOE sites via the GOCO Environmental Restoration Committee

  5. Implementation of the Westinghouse nuclear design system for incore fuel management analysis

    International Nuclear Information System (INIS)

    Hoskins, K.C.; Kichty, M.J.; Liu, Y.S.; Nguyen, T.Q.

    1990-01-01

    Development of the Westinghouse Advanced Nuclear Design System, which includes PHOENIX-P and ANC, has been continued to improve the efficiency, reliability, accuracy, and flexibility of models. The new codes ALPHA and PHIRE provide complete automation and interface functions for PHOENIX-P, ANC, and other codes. PHOENIX-P has been modified to generate data for ANC based on single or multi-assembly calculations. ANC has several enhancements, including improved pin power reconstruction, automated 2D model generation, and rod burnup prediction capability. The excellent performance of PHOENIX-P/ANC models is demonstrated by the results of over 30 models covering the range of Westinghouse designs. This Nuclear Design System is now the standard Westinghouse methodology for core design and analysis

  6. Seismic risk analysis for the Westinghouse Electric facility, Cheswick, Pennsylvania

    International Nuclear Information System (INIS)

    1977-01-01

    This report presents the results of a detailed seismic risk analysis of the Westinghouse Electric plutonium fuel development facility at Cheswick, Pennsylvania. This report focuses on earthquakes. The historical seismic record was established after a review of available literature, consultation with operators of local seismic arrays and examination of appropriate seismic data bases. Because of the aseismicity of the region around the site, an analysis different from the conventional closest approach in a tectonic province was adapted. Earthquakes as far from the site as 1,000 km were included, as were the possibility of earthquakes at the site. In addition, various uncertainties in the input were explicitly considered in the analysis. For example, allowance was made for both the uncertainty in predicting maximum possible earthquakes in the region and the effect of the dispersion of data about the best fit attenuation relation. The attenuation relationship is derived from two of the most recent, advanced studies relating earthquake intensity reports and acceleration. Results of the risk analysis, which include a Bayesian estimate of the uncertainties, are presented as return period accelerations. The best estimate curve indicates that the Westinghouse facility will experience 0.05 g every 220 years and 0.10 g every 1400 years. The accelerations are very insensitive to the details of the source region geometries or the historical earthquake statistics in each region and each of the source regions contributes almost equally to the cumulative risk at the site

  7. How Westinghouse is consolidating its international lead

    Energy Technology Data Exchange (ETDEWEB)

    1975-12-01

    The second of a series of profiles of major industrial groups in the world's nuclear industry, examines the attitudes and objectives of some of the executives now responsible for directing the widespread and complex international nuclear business of the Westinghouse Electric Corporation. Against the background of new management thinking in the group, the article discusses the significance of the emphasis on plant standardization of reliability, and on productivity in manufacturing.

  8. Human plan of capital of Westinghouse

    International Nuclear Information System (INIS)

    Alonso, B.; Gutierrez Elso, J. E.

    2008-01-01

    After three decades of nuclear standstill, the Nuclear Renaissance resulted in a changing environment, Nuclear Companies should prepare and adapt to different challenges: the fast growing of the organization, the loss of talent to other more attractive industrial fields and the transfer and management of knowledge to young engineers that have not participated in the building of nuclear plants. In this article different Westinghouse initiatives in this respect are commented. (Author)

  9. Leak rate analysis of the Westinghouse Reactor Coolant Pump

    International Nuclear Information System (INIS)

    Boardman, T.; Jeanmougin, N.; Lofaro, R.; Prevost, J.

    1985-07-01

    An independent analysis was performed by ETEC to determine what the seal leakage rates would be for the Westinghouse Reactor Coolant Pump (RCP) during a postulated station blackout resulting from loss of ac electric power. The object of the study was to determine leakage rates for the following conditions: Case 1: All three seals function. Case 2: No. 1 seal fails open while Nos. 2 and 3 seals function. Case 3: All three seals fail open. The ETEC analysis confirmed Westinghouse calculations on RCP seal performance for the conditions investigated. The leak rates predicted by ETEC were slightly lower than those predicted by Westinghouse for each of the three cases as summarized below. Case 1: ETEC predicted 19.6 gpm, Westinghouse predicted 21.1 gpm. Case 2: ETEC predicted 64.7 gpm, Westinghouse predicted 75.6 gpm. Case 3: ETEC predicted 422 gpm, Westinghouse predicted 480 gpm. 3 refs., 22 figs., 6 tabs

  10. Criticality safety training at Westinghouse Hanford Company

    International Nuclear Information System (INIS)

    Rogers, C.A.; Paglieri, J.N.

    1983-01-01

    In 1972 the Westinghouse Hanford Company (WHC) established a comprehensive program to certify personnel who handle fissionable materials. As the quantity of fissionable material handled at WHC has increased so has the scope of training to assure that all employes perform their work in a safe manner. This paper describes training for personnel engaged in fuel fabrication and handling activities. Most of this training is provided by the Fissionable Material Handlers Certification Program. This program meets or exceeds all DOE requirements for training and has been attended by more than 475 employes. Since the program was instituted, the rate of occurrence of criticality safety limit violations has decreased by 50%

  11. Westinghouse Water Reactor Divisions quality assurance plan

    International Nuclear Information System (INIS)

    1977-09-01

    The Quality Assurance Program used by Westinghouse Water Reactor Divisions is described. The purpose of the program is to assure that the design, materials, and workmanship on Nuclear Steam Supply System (NSSS) equipment meet applicable safety requirements, fulfill the requirements of the contracts with the applicants, and satisfy the applicable codes, standards, and regulatory requirements. This program satisfies the NRC Quality Assurance Criteria, 10CFR50 Appendix B, to the extent that these criteria apply to safety related NSSS equipment. Also, it follows the regulatory position provided in NRC regulatory guides and the requirements of ANSI Standard N45.2.12 as identified in this Topical Report

  12. Westinghouse Hanford Company environmental surveillance annual report

    International Nuclear Information System (INIS)

    Schmidt, J.W.; Johnson, A.R.; McKinney, S.M.; Perkins, C.J.; Webb, C.R.

    1992-07-01

    This document presents the results of near-facility operational environmental monitoring in 1991 of the 100, 200/600, and 300/400 Areas of the Hanford Site, in south-central Washington State, as performed by Westinghouse Hanford Company. These activities are conducted to assess and to control the impacts of operations on the workers and the local environment and to monitor diffuse sources. Surveillance activities include sampling and analyses of ambient air, surface water, groundwater, sediments, soil, and biota. Also, external radiation measurements and radiological surveys are taken at waste disposal sites, radiologically controlled areas, and roads

  13. Westinghouse GOCO conduct of casualty drills

    International Nuclear Information System (INIS)

    Ames, C.P.

    1996-02-01

    Purpose of this document is to provide Westinghouse Government Owned Contractor Operated (GOCO) Facilities with information that can be used to implement or improve drill programs. Elements of this guide are highly recommended for use when implementing a new drill program or when assessing an existing program. Casualty drills focus on response to abnormal conditions presenting a hazard to personnel, environment, or equipment; they are distinct from Emergency Response Exercises in which the training emphasis is on site, field office, and emergency management team interaction. The DOE documents which require team training and conducting drills in nuclear facilities and should be used as guidance in non-nuclear facilities are: DOE 5480.19 (Chapter 1 of Attachment I) and DOE 5480.20 (Chapter 1, paragraphs 7 a. and d. of continuing training). Casualty drills should be an integral part of the qualification and training program at every DOE facility

  14. Westinghouse Hanford Company waste minimization actions

    International Nuclear Information System (INIS)

    Greenhalgh, W.O.

    1988-09-01

    Companies that generate hazardous waste materials are now required by national regulations to establish a waste minimization program. Accordingly, in FY88 the Westinghouse Hanford Company formed a waste minimization team organization. The purpose of the team is to assist the company in its efforts to minimize the generation of waste, train personnel on waste minimization techniques, document successful waste minimization effects, track dollar savings realized, and to publicize and administer an employee incentive program. A number of significant actions have been successful, resulting in the savings of materials and dollars. The team itself has been successful in establishing some worthwhile minimization projects. This document briefly describes the waste minimization actions that have been successful to date. 2 refs., 26 figs., 3 tabs

  15. Helium leak testing the Westinghouse LCP coil

    International Nuclear Information System (INIS)

    Merritt, P.A.; Attaar, M.H.; Hordubay, T.D.

    1983-01-01

    The tests, equipment, and techniques used to check the Westinghouse LCP coil for coolant flow path integrity and helium leakage are unique in terms of test sensitivity and application. This paper will discuss the various types of helium leak testing done on the LCP coil as it enters different stages of manufacture. The emphasis will be on the degree of test sensitivity achieved under shop conditions, and what equipment, techniques and tooling are required to achieve this sensitivity (5.9 x 10 -8 scc/sec). Other topics that will be discussed are helium flow and pressure drop testing which is used to detect any restrictions in the flow paths, and the LCP final acceptance test which is the final leak test performed on the coil prior to its being sent for testing. The overall allowable leak rate for this coil is 5 x 10 -6 scc/sec. A general evaluation of helium leak testing experience are included

  16. Update of operations with Westinghouse steam generators

    International Nuclear Information System (INIS)

    Malinowski, D.D.; Fletcher, W.D.

    1978-01-01

    Westinghouse commercial steam generators in operation now number 112, of which 98 are tubed with Inconel 600, the remainder with stainless steel. The implementation of all volatile treatment (AVT) was reported. It was noted that several plants had exhibited some tube corrosion during their initial periods using AVT; this observation indicated that the transition from phosphate chemistry control to AVT may have been subject to certain residual effects due to incomplete removal of phosphated deposits. As inspection results from steam generators operated on AVT became more generally available with the passage of time, a pattern of results emerged that seemed to correlate with the operating experience with phosphate chemistry control. Specifically, all the plants that experienced corrosion problems had from 1 to 8 yr of operational history using phosphates, while those with less than a year's experience using phosphates tended to be less affected by corrosion problems

  17. Westinghouse Hanford Company Engineering Indoctrination Program

    International Nuclear Information System (INIS)

    Hull, K.J.

    1991-02-01

    Westinghouse Hanford Company has recognized that a learning curve exists in its engineering design programs. A one-year training program is under way to shorten this learning curve by introducing new engineers, both recent graduates and experienced new hires, to both company standards and intuitive engineering design processes. The participants are organized into multi-disciplined teams and assigned mentor engineers who assist them in completing a team project. Weekly sessions alternate between information presentations and time to work on team design projects. The presentations include information that is applicable to the current phase of the design project as well as other items of interest, such as site tours, creative thinking, and team brainstorming techniques. 1 fig

  18. Human plan of capital of Westinghouse; Plan de capital humano de Westinghouse

    Energy Technology Data Exchange (ETDEWEB)

    Alonso, B.; Gutierrez Elso, J. E.

    2008-07-01

    After three decades of nuclear standstill, the Nuclear Renaissance resulted in a changing environment, Nuclear Companies should prepare and adapt to different challenges: the fast growing of the organization, the loss of talent to other more attractive industrial fields and the transfer and management of knowledge to young engineers that have not participated in the building of nuclear plants. In this article different Westinghouse initiatives in this respect are commented. (Author)

  19. Westinghouse experience in the transfer of nuclear technology

    International Nuclear Information System (INIS)

    Simpson, J.W.

    1977-01-01

    Westinghouse experience with transfer of technical information is two-sided. First is our experience in learning, and the second is our experience in teaching others. Westinghouse conducts a special school to which government, academic and industry people are invited. There are many problems involved in all technology transfers; these include: keeping information current, making certain changes are compatible with the supplier's manufacturing capability and also suitable to the receiver, patent right and proprietary information. The building, testing and maintenance of the unit on the line - and then a succession of its sister plant is the basis for the Westinghouse leadership

  20. Westinghouse says cartel rigged U.S. uranium market

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    On Oct. 15, 1976, Westinghouse filed a complaint in Federal court in Chicago charging that 29 U.S. and foreign uranium producers damaged Westinghouse by illegally rigging the uranium market; they also link the Atomic Industrial Forum to the U.S. activities of this cartel. Background information is presented for the charge, which has become the focal point of Westinghouse's defense against the uranium supply breach of contract suits filed against the firm by 27 electric utilities (3 filed in county court in Pittsburgh, 24 jointly in Federal court in Virginia). Westinghouse attorneys say that most of the evidence they have shows the existence of a cartel in the past, but they hope to show it is still operating in the U.S

  1. The investigation of Passive Accident Mitigation Scheme for advanced PWR NPP

    International Nuclear Information System (INIS)

    Shi, Er-bing; Fang, Cheng-yue; Wang, Chang; Xia, Geng-lei; Zhao, Cui-na

    2015-01-01

    Highlights: • We put forward a new PAMS and analyze its operation characteristics under SBO. • We conduct comparative analysis between PAMS and Traditional Secondary Side PHRS. • The PAMS could cope with SBO accident and maintain the plant in safe conditions. • PAMS could decrease heat removal capacity of PHRS. • PAMS has advantage in reducing cooling rate and PCCT temperature rising amplitude. - Abstract: To enhance inherent safety features of nuclear power plant, the advanced pressurized water reactors implement a series of passive safety systems. This paper puts forward and designs a new Passive Accident Mitigation Scheme (PAMS) to remove residual heat, which consists of two parts: the first part is Passive Auxiliary Feedwater System (PAFS), and the other part is Passive Heat Removal System (PHRS). This paper takes the Westinghouse-designed Advanced Passive PWR (AP1000) as research object and analyzes the operation characteristics of PAMS to cope with the Station Blackout Accident (SBO) by using RELAP5 code. Moreover, the comparative analysis is also conducted between PAMS and Traditional Secondary Circuit PHRS to derive the advantages of PAMS. The results show that the designed scheme can remove core residual heat significantly and maintain the plant in safe conditions; the first part of PAMS would stop after 120 min and the second part has to come into use simultaneously; the low pressurizer (PZR) pressure signal would be generated 109 min later caused by coolant volume shrinkage, which would actuate the Passive Safety Injection System (PSIS) to recovery the water level of pressurizer; the flow instability phenomenon would occur and last 21 min after the PHRS start-up; according to the comparative analysis, the coolant average temperature gradient and the Passive Condensate Cooling Tank (PCCT) water temperature rising amplitude of PAMS are lower than those of Traditional Secondary Circuit PHRS

  2. Overview on advanced nuclear reactors: research and deployment in the United States

    International Nuclear Information System (INIS)

    Sandell, L.; Rohrer, S.

    2004-01-01

    For the United States of America, the electricity requirement is expected to continue to rise at rates of approximately 1.8% over the next few years. This means that some 300,000 MW of additional generating capacity need to be made available by 2025. The Energy Policy Act of 2003 is to minimize this expected future growth of electricity consumption and promote research in favor of a diversified energy mix. As a consequence, the U.S. Senate and the House of Representatives passed legislation on electricity generation, on the promotion of, and research into, specific energy sources, and on energy conservation. Currently, coal-fired power plants contribute the largest share to the overall generating capacity. Considerable additions to the generating capacity have been made in the past ten years in gas-fired plants. In the light of the high present gas prices and market volatilities, the construction of new coal-fired power plants is currently under discussion. 103 out of the 436 nuclear power plants at present in operation worldwide are located in the United States. They represent by far the largest share of emission-free generating capacity in the United States. Considerable capacities have been added over the past few years by, up to now, 99 power increases by 0.4 to 17.8%. The Nuclear Power 2010 Program is a joint initiative by the government and industry seeking to further develop advanced nuclear power plant technologies and elaborate a new licensing procedure for nuclear power plants. The proposed licensing procedure and the Westinghouse AP1000, General Electric ESBWR, and AECL ACR-700 advanced reactor lines are presented. (orig.)

  3. Standardized Technical Specifications for Westinghouse PWRs

    International Nuclear Information System (INIS)

    1978-01-01

    This Standard Technical Specification (STS) has been structured for the broadest possible use on Westinghouse plants currently being reviewed for an Operating License. Accordingly, the document contains specifications applicable to plants (1) with either 3 or 4 loops and (2) with and without loop stop valves. In addition, four separate and discrete containment specification sections are provided for each of the following containment types: Atmospheric, Ice Condenser, Sub-Atmospheric, and Dual. Optional specifications are provided for those features and systems which may be included in individual plant designs but are not generic in their scope of application. Alternate specifications are provided in a limited number of cases to cover situations where alternate specification requirements are necessary on a generic basis because of design differences. The format of the STS addresses the categories required by 10 CFR 50 and consists of six sections covering the areas of: Definitions, Safety Limits and Limiting Safety System Settings, Limiting Conditions for Operation, Surveillance Requirements, Design Features, and Administrative Controls

  4. Westinghouse waste simulation and optimization software tool

    International Nuclear Information System (INIS)

    Mennicken, Kim; Aign, Jorg

    2013-01-01

    Applications for dynamic simulation can be found in virtually all areas of process engineering. The tangible benefits of using dynamic simulation can be seen in tighter design, smoother start-ups and optimized operation. Thus, proper implementation of dynamic simulation can deliver substantial benefits. These benefits are typically derived from improved process understanding. Simulation gives confidence in evidence based decisions and enables users to try out lots of 'what if' scenarios until one is sure that a decision is the right one. In radioactive waste treatment tasks different kinds of waste with different volumes and properties have to be treated, e.g. from NPP operation or D and D activities. Finding a commercially and technically optimized waste treatment concept is a time consuming and difficult task. The Westinghouse Waste Simulation and Optimization Software Tool will enable the user to quickly generate reliable simulation models of various process applications based on equipment modules. These modules can be built with ease and be integrated into the simulation model. This capability ensures that this tool is applicable to typical waste treatment tasks. The identified waste streams and the selected treatment methods are the basis of the simulation and optimization software. After implementing suitable equipment data into the model, process requirements and waste treatment data are fed into the simulation to finally generate primary simulation results. A sensitivity analysis of automated optimization features of the software generates the lowest possible lifecycle cost for the simulated waste stream. In combination with proven waste management equipments and integrated waste management solutions, this tool provides reliable qualitative results that lead to an effective planning and minimizes the total project planning risk of any waste management activity. It is thus the ideal tool for designing a waste treatment facility in an optimum manner

  5. Westinghouse waste simulation and optimization software tool

    Energy Technology Data Exchange (ETDEWEB)

    Mennicken, Kim; Aign, Jorg [Westinghouse Electric Germany GmbH, Hamburg (Germany)

    2013-07-01

    Applications for dynamic simulation can be found in virtually all areas of process engineering. The tangible benefits of using dynamic simulation can be seen in tighter design, smoother start-ups and optimized operation. Thus, proper implementation of dynamic simulation can deliver substantial benefits. These benefits are typically derived from improved process understanding. Simulation gives confidence in evidence based decisions and enables users to try out lots of 'what if' scenarios until one is sure that a decision is the right one. In radioactive waste treatment tasks different kinds of waste with different volumes and properties have to be treated, e.g. from NPP operation or D and D activities. Finding a commercially and technically optimized waste treatment concept is a time consuming and difficult task. The Westinghouse Waste Simulation and Optimization Software Tool will enable the user to quickly generate reliable simulation models of various process applications based on equipment modules. These modules can be built with ease and be integrated into the simulation model. This capability ensures that this tool is applicable to typical waste treatment tasks. The identified waste streams and the selected treatment methods are the basis of the simulation and optimization software. After implementing suitable equipment data into the model, process requirements and waste treatment data are fed into the simulation to finally generate primary simulation results. A sensitivity analysis of automated optimization features of the software generates the lowest possible lifecycle cost for the simulated waste stream. In combination with proven waste management equipments and integrated waste management solutions, this tool provides reliable qualitative results that lead to an effective planning and minimizes the total project planning risk of any waste management activity. It is thus the ideal tool for designing a waste treatment facility in an optimum manner

  6. Implementation of the Westinghouse WRB-2 CHF correlation in VIPRE

    International Nuclear Information System (INIS)

    Klasmier, L.K.; Haksoo Kim

    1992-01-01

    As part of the reload transient and thermal-hydraulic methods development effort within Commonwealth Edison Company (CECo), the WRB-2 critical heat flux (CHF) correlation has been implemented into the VIPRE-01 thermal-hydraulic analysis code to support Westinghouse 17X17 Vantage 5 fuel. CECo is in the process of switching from Westinghouse optimized fuel assembly (OFA) fuel to Vantage 5 fuel at CECo's six pressurized water reactors. CECo performs the neutronic portion of the reload analysis using Westinghouse's ANC/PHOENIX. The transient and thermal-hydraulic analysis will be performed using the RETRAN and VIPRE codes once the Nuclear Regulatory Commission has completed their review of CECo methodology. Previously, CECo had implemented and received NRC approval to use the Westinghouse WRB-1 CHF correlation in the VIPRE-01 code to support 15X15 and 17X17 OFA fuel designs. Since the WRB-1 CHF correlation is not applicable for 17X17 Vantage 5 fuel, it was necessary to implement the WRB-2 CHF correlation in the VIPRE code. The WRB-2 correlation was developed by Westinghouse using a database applicable to 17X17 OFA and Vantage 5 fuel and the THINC thermal-hydraulic analysis code. At CECo, the WRB-2 correlation had been implemented into VIPRE-01/MOD-02. The results produced at CECo have been statistically compared to those produced by Westinghouse. Owen's method was used to determine the VIPRE/WRB-02 thermal limit. The thermal limit for 17X17 OFA and Vantage 5 fuel use in VIPRE/WRB-2 is in excellent agreement with the value calculated by Westinghouse using THINC/WRB-2

  7. Performance of the Westinghouse WWER-1000 fuel design

    International Nuclear Information System (INIS)

    Hoglund, J.; Riznychenko, O.; Latorre, R.; Lashevych, P.

    2011-01-01

    In 2005 six (6) Westinghouse WWER-1000 Lead Test Assemblies (LTAs) were loaded in the South Ukraine Unit 3. This design has demonstrated full compatibility with resident fuel designs and all associated fuel handling and reactor components. Operations have further demonstrated adequacy of performance margins and the reliability requirements for multiple cycles of operation. The LTA's have now been discharged after completing the planned four cycles of operation and having reached an average assembly burnup in excess of 43 MWd/kgU. Post Irradiation Examinations were performed after completion of each cycle. The final LTA inspection program at end of Cycle 20 in 2010 yielded satisfactory results on all counts, and it was concluded that the 6 Westinghouse LTA's performed as expected during their operational regimes. Very good performance was demonstrated in the WWER-1000 reactor environment for the Zr-1%Nb as grid material, and ZIRLO fuel cladding and structural components. Control Rod Assemblies drop times and drag forces were all within the accepted values. The LTA program demonstrated that this fuel design is suitable for full core applications. However, the topic of fuel assembly distortion resistance was re-visited and Westinghouse therefore considered operational experience and design features from multiple development programs to enhance the basic Westinghouse WWER-1000 fuel design for Ukrainian reactors. The design now includes features that further mitigate assembly bow while at the same time improving the fuel cycle economy. This paper describes briefly the development of the Westinghouse WWER-1000 fuel design and how test results and operational experiences from multiple sources have been utilized to produce a most suitable fuel design. Early in 2011 a full region of the Westinghouse WWER-1000 design completed another full cycle of operation at South Ukraine Unit 3, all with excellent results. All 42 fuel assemblies were examined for visible damage or non

  8. Westinghouse fuel manufacturing systems: a step change in performance improvements

    International Nuclear Information System (INIS)

    Mutyala, Meena

    2009-01-01

    Today's competitive electrical generation industry demands that nuclear power plant operators minimize total operating costs, including fuel cycle cost while maintaining flawless fuel performance. The mission of Westinghouse Nuclear Fuel is to be the industry's most responsive supplier of flawless, value added fuel products and services, as judged by our customers. As nuclear is fast becoming the choice of many countries, existing manufacturing plants and facilities are once again running at full capacity. In this context Westinghouse Nuclear Fuel is committed to deliver a step change in performance improvement worldwide through its manufacturing operations by the introduction of a set of fundamentals collectively named the 'Westinghouse Fuel Manufacturing System' (WFMS), whose key principles are discussed in this paper. (author)

  9. Exporting advanced service tech to Europe

    International Nuclear Information System (INIS)

    Naredo, F.P.

    1991-01-01

    The concept of Westinghouse's European Service Center (ESC), located at Nivelles, Belgium, to bring advanced services and maintenance technologies to European customers is described. Laser-welded sleeving to repair degraded steam generator tubes employing pulsed Neodymium Yttrium Aluminium Garnet (Nd:YAG) laser, which Westinghouse has developed for 22mm tubing is mentioned. U-bend heat treatment (UBHT) technology was also brought to Europe by Westinghouse

  10. Westinghouse accident tolerant fuel program. Current results and future plans

    Energy Technology Data Exchange (ETDEWEB)

    Ray, Sumit; Xu, Peng; Lahoda, Edward; Hallstadius, Lars; Boylan, Frank [Westinghouse Electric Company LLC, Hopkins, SC (United States)

    2016-07-15

    This paper discusses the current status, results from initial tests, as well as the future direction of the Westinghouse's Accident Tolerant Fuel (ATF) program. The current preliminary testing is addressed that is being performed on these samples at the Massachusetts Institute of Technology (MIT) test reactor, initial results from these tests, as well as the technical learning from these test results. In the Westinghouse ATF approach, higher density pellets play a significant role in the development of an integrated fuel system.

  11. Standard Technical Specifications for Westinghouse pressurized water reactors

    International Nuclear Information System (INIS)

    Virgilio, M.J.

    1980-09-01

    The Standard Technical Specifications for Westinghouse Pressurized Water Reactors (W-STS) is a generic document prepared by the U.S. NRC for use in the licensing process of current Westinghouse Pressurized Water Reactors. The W-STS sets forth the Limits, Operating Conditions and other requirements applicable to nuclear reactor facility operation as set forth in by Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. This document is revised periodically to reflect current licensing requirements

  12. The characteristics of the Westinghouse accident procedures and the main differences with SOP

    International Nuclear Information System (INIS)

    Hu Yan; Gan Peijiang; Sun Chen

    2014-01-01

    In this note, the Westinghouse operation file system is summarized. The structures of procedures, design methods, implementation logics of the Westinghouse accident procedures are discussed. And compared with the SOP principles, the main differences are clarified. (authors)

  13. Overview of expert systems applications in Westinghouse Nuclear Fuel Activities

    International Nuclear Information System (INIS)

    Leech, W.J.

    1989-01-01

    Expert system applications have been introduced in several nuclear fuel activities, including engineering and manufacturing. This technology has been successfully implemented on the manufacturing floors to provide on-line process control at zirconium tubing and fuel fabrication plants. This paper provides an overview of current applications at Westinghouse with respect to fuel fabrication, zirconium tubing, zirconium production, and core design

  14. An Overview of Westinghouse Realistic Large Break LOCA Evaluation Model

    Directory of Open Access Journals (Sweden)

    Cesare Frepoli

    2008-01-01

    Full Text Available Since the 1988 amendment of the 10 CFR 50.46 rule in 1988, Westinghouse has been developing and applying realistic or best-estimate methods to perform LOCA safety analyses. A realistic analysis requires the execution of various realistic LOCA transient simulations where the effect of both model and input uncertainties are ranged and propagated throughout the transients. The outcome is typically a range of results with associated probabilities. The thermal/hydraulic code is the engine of the methodology but a procedure is developed to assess the code and determine its biases and uncertainties. In addition, inputs to the simulation are also affected by uncertainty and these uncertainties are incorporated into the process. Several approaches have been proposed and applied in the industry in the framework of best-estimate methods. Most of the implementations, including Westinghouse, follow the Code Scaling, Applicability and Uncertainty (CSAU methodology. Westinghouse methodology is based on the use of the WCOBRA/TRAC thermal-hydraulic code. The paper starts with an overview of the regulations and its interpretation in the context of realistic analysis. The CSAU roadmap is reviewed in the context of its implementation in the Westinghouse evaluation model. An overview of the code (WCOBRA/TRAC and methodology is provided. Finally, the recent evolution to nonparametric statistics in the current edition of the W methodology is discussed. Sample results of a typical large break LOCA analysis for a PWR are provided.

  15. Factory Acceptance Test Procedure Westinghouse 100 ton Hydraulic Trailer

    International Nuclear Information System (INIS)

    Aftanas, B.L.

    1994-01-01

    This Factory Acceptance Test Procedure (FAT) is for the Westinghouse 100 Ton Hydraulic Trailer. The trailer will be used for the removal of the 101-SY pump. This procedure includes: safety check and safety procedures; pre-operation check out; startup; leveling trailer; functional/proofload test; proofload testing; and rolling load test

  16. 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.

  17. 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

  18. Westinghouse, DOE see apples, oranges in IG staffing report

    International Nuclear Information System (INIS)

    Lobsenz, G.

    1994-01-01

    The operator of the Energy Department's Savannah River weapons plant has at least 1,800 more employees than it needs, and could save $400 million over a five-year period by cutting its staff accordingly, a DOE inspector general study says. Most of the boat - 1,206 employees - was attributed to excessive numbers of managers, with the inspector general concluding that Westinghouse Savannah River Co. had roughly twice as many layers of management than two other DOE weapons contractors. The study also concluded that Westinghouse in fiscal year 1992 significantly understated its actual staffing levels in reports to DOE, failing to disclose 1,765 full-time employees or the equivalent hours worked. Through such underreporting Westinghouse was able to open-quotes circumvent staffing ceilings established by the department,close quotes the study added. Overall, DOE Inspector General John Layton said Westinghouse's staff levels substantially exceeded those needed for efficient operation of the South Carolina nuclear weapons facility. Layton based his analysis on efficiency standards attained by other DOE weapons plant contractors, such as Martin Marietta Energy Systems at DOE's Oak Ridge, Tenn., plant and EG ampersand G Rocky Flats, as well as widely utilized worker performance requirements used by the Navy and private sector companies that perform work similar to that done at Savannah River

  19. Root cause of incomplete control rod insertions at Westinghouse reactors

    International Nuclear Information System (INIS)

    Ray, S.

    1997-01-01

    Within the past year, incomplete RCCA insertions have been observed on high burnup fuel assemblies at two Westinghouse PWRs. Initial tests at the Wolf Creek site indicated that the direct cause of the incomplete insertions observed at Wolf Creek was excessive fuel assembly thimble tube distortion. Westinghouse committed to the NRC to perform a root cause analysis by the end of August, 1996. The root cause analysis process used by Westinghouse included testing at ten sites to obtain drag, growth and other characteristics of high burnup fuel assemblies. It also included testing at the Westinghouse hot cell of two of the Wolf Creek incomplete insertion assemblies. A mechanical model was developed to calculate the response of fuel assemblies when subjected to compressive loads. Detailed manufacturing reviews were conducted to determine if this was a manufacturing related issue. In addition, a review of available worldwide experience was performed. Based on the above, it was concluded that the thimble tube distortion observed on the Wolf Creek incomplete insertion assemblies was caused by unusual fuel assembly growth over and above what would typically be expected as a result of irradiation exposure. It was determined that the unusual growth component is a combination of growth due to oxide accumulation and accelerated growth, and would only be expected in high temperature plants on fuel assemblies that see long residence times and high power duties

  20. Perspective of the Westinghouse steam generator secondary side maintenance approach

    Energy Technology Data Exchange (ETDEWEB)

    Ramaley, D. [Westinghouse Electric Company LLC, Cranberry Township, Pennsylvania (United States)

    2012-07-01

    Historically, Westinghouse had developed a set of steam generator secondary maintenance guidelines focused around performing recurring activities each outage without direct regards to the age, deposit loading, operational status, or corrosion status of the steam generator. Through the evolution of steam generator design and steam generator condition data, Westinghouse now uses a proactive assessment and planning approach for utilities. Westinghouse works with utilities to develop steam generator secondary maintenance plans for long term steam generator viability. Westinghouse has developed a portfolio of products to allow utilities to optimize steam generator operability and develop programs aimed at maintaining the steam generator secondary side in a favorable condition for successful long term operation. Judicious use of the means available for program development should allow for corrosion free operation, long term full power operation at optimum thermal efficiency, and leveling of outage expenditures over a long period of time. This paper will review the following required elements for an effective steam generator secondary side strategy: • Assessment: In order to develop an appropriate maintenance strategy, actions must be taken to obtain an accurate picture of the SG secondary side condition. • Forecasting: Using available data predictions are developed for future steam generator conditions and required maintenance actions. • Action: Cost effective engineering and maintenance actions must be completed at the appropriate time as designated by the plan. • Evaluation of Results: Following execution of maintenance tactics, it is necessary to revise strategy and develop technology enhancements as appropriate. (author)

  1. Initial performance assessment of the Westinghouse AP600 containment design and related safety issues

    International Nuclear Information System (INIS)

    Nicolette, V.F.; Washington, K.E.; Tills, J.L.

    1991-01-01

    This work summarizes the Westinghouse AP600 advanced reactor design assessment calculations performed to date with the CONTAIN code. Correlations for modeling the important heat transfer phenomena are discussed as well. A CONTAIN model of the AP600 was constructed for design basis accident (DBA) calculations. Insights gained from modeling of the smaller-scale Westinghouse Integral Test Facility were incorporated in the development of the AP600 model. The results of the DBA calculations are compared to the results of other researchers to serve as a point of reference for future severe accident calculations. The CONTAIN calculations are reviewed to examine several parameters/phenomena of interest. The results of the calculations are also used to identify limitations of the CONTAIN code regarding application to advanced reactor containment designs. The most recent heat transfer correlations available in the literature are assessed for use in the flow regimes and geometries applicable to the AP600. Use of one of these correlations in CONTAIN may allow for a more accurate assessment of the AP600

  2. WGOTHIC analysis of AP1000 passive containment cooling water

    International Nuclear Information System (INIS)

    Ye Cheng; Wang Yong; Zheng Mingguang; Wang Guodong; Zhang Di; Ni Chenxiao; Wang Minglu

    2013-01-01

    The WGOTHIC code was used to analyze the influence of the containment cooling water inventory to containment safety for different cases. The results show that if passive containment cooling system fails, the pressure in containment is beyond design limit after 1000 s; if cooling water can't be supplied after 72 h, the pressure in containment is beyond design limit after 0.9 d; if cooling water can't be supplied after 19.6 d, the pressure in containment is beyond design limit but less than the breakdown pressure; if cooling water is supplied for 30 d, the air cooling can remove the decay heat without any aid. It is a reference for making emergency plan and improving containment design. (authors)

  3. 76 FR 82079 - AP1000 Design Certification Amendment

    Science.gov (United States)

    2011-12-30

    ... generators, the refueling water storage tank, and various equipment for power generation, refueling, and... in settling of any resultant debris) to facilitate heat transfer to the containment vessel and for...). The source of water for the evaporative cooling is the passive containment cooling water storage tank...

  4. 76 FR 10269 - AP1000 Design Certification Amendment

    Science.gov (United States)

    2011-02-24

    ... (Chapter 3), instrumentation and control (I&C) systems (Chapter 7) and human factors engineering (HFE... proposed such changes. Once the NRC was satisfied with these DCD markups, they were documented in the... FSER. This chapter indicates which areas of the DCD are affected by each design change and the letters...

  5. Overview of the Westinghouse Small Modular Reactor building layout

    Energy Technology Data Exchange (ETDEWEB)

    Cronje, J. M. [Westinghouse Electric Company LLC, Centurion (South Africa); Van Wyk, J. J.; Memmott, M. J. [Westinghouse Electric Company LLC, Cranberry Township, PA (United States)

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the third in a series of four papers, which describe the design and functionality of the Westinghouse SMR. It focuses in particular upon the plant building layout and modular design of the Westinghouse SMR. In the development of small modular reactors, the building layout is an area where the safety of the plant can be improved by applying new design approaches. This paper will present an overview of the Westinghouse SMR building layout and indicate how the design features improve the safety and robustness of the plant. The Westinghouse SMR is designed with no shared systems between individual reactor units. The main buildings inside the security fence are the nuclear island, the rad-waste building, the annex building, and the turbine building. All safety related equipment is located in the nuclear island, which is a seismic class 1 building. To further enhance the safety and robustness of the design, the reactor, containment, and most of the safety related equipment are located below grade on the nuclear island. This reduces the possibility of severe damage from external threats or natural disasters. Two safety related ultimate heat sink (UHS) water tanks that are used for decay heat removal are located above grade, but are redundant and physically separated as far as possible for improved safety. The reactor and containment vessel are located below grade in the center of the nuclear island. The rad-waste and other radioactive systems are located on the bottom floors to limit the radiation exposure to personnel. The Westinghouse SMR safety trains are completely separated into four unconnected quadrants of the building, with access between quadrants only allowed

  6. Drop testing of the Westinghouse fresh nuclear fuel package

    International Nuclear Information System (INIS)

    Shappert, L.B.; Sanders, C.F.

    1993-01-01

    The Westinghouse Columbia Fuel Fabrication Facility has decided to develop and certify a new fresh fuel package design (type A, fissile) that has the capability to transport more highly enriched fuel than was previously possible. A prototype package was tested in support of the Safety Analysis Report of the Packaging (SARP). This paper provides detailed information on the tests and test results. A first prototype test was carried out at the STF, and the design did not give the safety margin that Westinghouse wanted for their containers. The data from the test were used to redesign the connection between the clamping frame and the pressure pad, and the tests were reinitiated. Three packages were then tested at the STF. All packages met the acceptance criteria and acceleration information was obtained that provided an indication of the behavior of the cradle and strongback which holds the fuel assemblies and nuclear poison in place. (J.P.N.)

  7. Chemical-cleaning process evaluation: Westinghouse steam generators. Final report

    International Nuclear Information System (INIS)

    Cleary, W.F.; Gockley, G.B.

    1983-04-01

    The Steam Generator Owners Group (SGOG)/Electric Power Research Institute (EPRI) Steam Generator Secondary Side Chemical Cleaning Program, under develpment since 1978, has resulted in a generic process for the removal of accumulated corrosion products and tube deposits in the tube support plate crevices. The SGOG/EPRI Project S150-3 was established to obtain an evaluation of the generic process in regard to its applicability to Westinghouse steam generators. The results of the evaluation form the basis for recommendations for transferring the generic process to a plant specific application and identify chemical cleaning corrosion guidelines for the materials in Westinghouse Steam Generators. The results of the evaluation, recommendations for plant-specific applications and corrosion guidelines for chemical cleaning are presented in this report

  8. Drop testing of the Westinghouse fresh nuclear fuel package

    International Nuclear Information System (INIS)

    Shappert, L.B.; Sanders, C.F.

    1992-01-01

    In recent years, the Westinghouse Columbia Fuel Fabrication Facility has been faced with increasing pressure from utilities that wished to take the fuel in their nuclear power plants to higher burnups. To help accommodate this trend, Westinghouse has determined that it needs the ability to increase the enrichment of the fresh fuel it delivers to its customers. One critical step in this process is to certify a new (Type A, fissile) fresh fuel package design that has the capability to transport fuel with a higher enrichment than was previously available. A prototype package was tested in support of the Safety Analysis Report of the Packaging. This paper provides detailed information on those tests and their results

  9. Current status of Westinghouse tubular solid oxide fuel cell program

    Energy Technology Data Exchange (ETDEWEB)

    Parker, W.G. [Westinghouse Science and Technology Center, Pittsburgh, PA (United States)

    1996-04-01

    In the last ten years the solid oxide fuel cell (SOFC) development program at Westinghouse has evolved from a focus on basic material science to the engineering of fully integrated electric power systems. Our endurance for this cell is 5 to 10 years. To date we have successfully operated at power for over six years. For power plants it is our goal to have operated before the end of this decade a MW class power plant. Progress toward these goals is described.

  10. Simulator testing of the Westinghouse aware alarm management system

    Energy Technology Data Exchange (ETDEWEB)

    Carrera, J P; Easter, J R; Roth, E M [Westinghouse Electric Corp., Pittsburgh, PA (United States)

    1997-09-01

    Over the last year, Westinghouse engineers and operators from the Beznau nuclear power station (KKB), owned by the Nordostschweizerische Krafwerke AG of Baden, Switzerland, have been installing and testing the Westinghouse AWARE Alarm Management System in Beznau/SNUPPS operator training simulator, owned and operated by the Westinghouse Electric Corp., in Waltz Mill, PA, USA. The testing has focused primarily on validating the trigger logic data base and on familiarizing the utility`s training department with the operation of the system in a real-time environment. Some of the tests have included plant process scenarios in which the computerized Emergency Procedures were available and used through the COMPRO (COMputerized PROcedures) System in conjunction with the AWARE System. While the results to date are qualitative from the perspective of system performance and improvement in message presentation, the tests have generally confirmed the expectations of the design. There is a large reduction in the number of messages that the control room staff must deal with during major process abnormalities, yet at times of relative minor disturbances, some additional messages are available which add clarification, e.g., ``Pump Trouble`` messages. The ``flow`` of an abnormality as it progresses from one part of the plant`s processes to another is quite visible. Timing of the messages and the lack of message avalanching is proving to give the operators additional time to respond to messages. Generally, the anxiety level to ``do something`` immediately upon a reactor trip appears to be reduced. (author). 8 refs.

  11. Simulator testing of the Westinghouse aware alarm management system

    International Nuclear Information System (INIS)

    Carrera, J.P.; Easter, J.R.; Roth, E.M.

    1997-01-01

    Over the last year, Westinghouse engineers and operators from the Beznau nuclear power station (KKB), owned by the Nordostschweizerische Krafwerke AG of Baden, Switzerland, have been installing and testing the Westinghouse AWARE Alarm Management System in Beznau/SNUPPS operator training simulator, owned and operated by the Westinghouse Electric Corp., in Waltz Mill, PA, USA. The testing has focused primarily on validating the trigger logic data base and on familiarizing the utility's training department with the operation of the system in a real-time environment. Some of the tests have included plant process scenarios in which the computerized Emergency Procedures were available and used through the COMPRO (COMputerized PROcedures) System in conjunction with the AWARE System. While the results to date are qualitative from the perspective of system performance and improvement in message presentation, the tests have generally confirmed the expectations of the design. There is a large reduction in the number of messages that the control room staff must deal with during major process abnormalities, yet at times of relative minor disturbances, some additional messages are available which add clarification, e.g., ''Pump Trouble'' messages. The ''flow'' of an abnormality as it progresses from one part of the plant's processes to another is quite visible. Timing of the messages and the lack of message avalanching is proving to give the operators additional time to respond to messages. Generally, the anxiety level to ''do something'' immediately upon a reactor trip appears to be reduced. (author). 8 refs

  12. Westinghouse experience in using mechanical cutting for reactor vessel internals segmentation

    International Nuclear Information System (INIS)

    Boucau, Joseph; Fallstroem, Stefan; Segerud, Per; Kreitman, Paul J.

    2010-01-01

    Some commercial nuclear power plants have been permanently shut down to date and decommissioned using dismantling methods. Other operating plants have decided to undergo an upgrade process that includes replacement of reactor internals. In both cases, there is a need to perform a segmentation of the reactor vessel internals with proven methods for long term waste disposal. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques. Mechanical cutting has been used by Westinghouse since 1999 for both PWRs and BWRs and its process has been continuously improved over the years. Detailed planning is essential to a successful project, and typically a 'Segmentation and Packaging Plan' is prepared to document the effort. The usual method is to start at the end of the process, by evaluating the waste disposal requirements imposed by the waste disposal agency, what type and size of containers are available for the different disposal options, and working backwards to select the best cutting tools and finally the cut geometry required. These plans are made utilizing advanced 3-D CAD software to model the process. Another area where the modelling has proven invaluable is in determining the logistics of component placement and movement in the reactor cavity, which is typically very congested when all the internals are out of the reactor vessel in various stages of segmentation. The main objective of the segmentation and packaging plan is to determine the strategy for separating the highly activated components from the less activated material, so that they can be disposed of in the most cost effective manner. Usually, highly activated components cannot be shipped off-site, so they must be packaged such that they can be dry stored with the spent fuel in an Independent Spent Fuel Storage Installation (ISFSI). Less activated components can be shipped to an off-site disposal site depending on space availability. Several of the

  13. The Westinghouse Hanford Company Operational Environmental Monitoring Program CY-93

    International Nuclear Information System (INIS)

    Schmidt, J.W.

    1993-10-01

    The Operational Environmental Monitoring Program (OEMP) provides facility-specific environmental monitoring to protect the environment adjacent to facilities under the responsibility of Westinghouse Hanford Company (WHC) and assure compliance with WHC requirements and local, state, and federal environmental regulations. The objectives of the OEMP are to evaluate: compliance with federal (DOE, EPA), state, and internal WHC environmental radiation protection requirements and guides; performance of radioactive waste confinement systems; and trends of radioactive materials in the environment at and adjacent to nuclear facilities and waste disposal sites. This paper identifies the monitoring responsibilities and current program status for each area of responsibility

  14. Westinghouse integrated cementation facility. Smart process automation minimizing secondary waste

    International Nuclear Information System (INIS)

    Fehrmann, H.; Jacobs, T.; Aign, J.

    2015-01-01

    The Westinghouse Cementation Facility described in this paper is an example for a typical standardized turnkey project in the area of waste management. The facility is able to handle NPP waste such as evaporator concentrates, spent resins and filter cartridges. The facility scope covers all equipment required for a fully integrated system including all required auxiliary equipment for hydraulic, pneumatic and electric control system. The control system is based on actual PLC technology and the process is highly automated. The equipment is designed to be remotely operated, under radiation exposure conditions. 4 cementation facilities have been built for new CPR-1000 nuclear power stations in China

  15. The Westinghouse Series 1000 Mobile Phone: Technology and applications

    Science.gov (United States)

    Connelly, Brian

    1993-01-01

    Mobile satellite communications will be popularized by the North American Mobile Satellite (MSAT) system. The success of the overall system is dependent upon the quality of the mobile units. Westinghouse is designing our unit, the Series 1000 Mobile Phone, with the user in mind. The architecture and technology aim at providing optimum performance at a low per unit cost. The features and functions of the Series 1000 Mobile Phone have been defined by potential MSAT users. The latter portion of this paper deals with who those users may be.

  16. Westinghouse use of artificial intelligence in signal interpretation

    International Nuclear Information System (INIS)

    Mark, R.H.

    1984-01-01

    This paper discusses Westinghouse's use of artificial intelligence to assist inspectors who routinely monitor the thousands of tubes in nuclear steam generators. Using the AI technology has made the inspection process easier to learn and to apply. The system uses pattern recognition to identify off-normal conditions. As part of the in-service inspection program for nuclear power reactors, utilities make a practice of inspecting the condition of the large heat exchangers that produce the steam that turns the electric turbine generator. The same data are presented for inspection using form, motion, and color to call attention to off-normal signal patterns

  17. Piping benchmark problems for the Westinghouse AP600 Standardized Plant

    International Nuclear Information System (INIS)

    Bezler, P.; DeGrassi, G.; Braverman, J.; Wang, Y.K.

    1997-01-01

    To satisfy the need for verification of the computer programs and modeling techniques that will be used to perform the final piping analyses for the Westinghouse AP600 Standardized Plant, three benchmark problems were developed. The problems are representative piping systems subjected to representative dynamic loads with solutions developed using the methods being proposed for analysis for the AP600 standard design. It will be required that the combined license licensees demonstrate that their solutions to these problems are in agreement with the benchmark problem set

  18. Westinghouse Hanford Company risk management strategy for retired surplus facilities

    International Nuclear Information System (INIS)

    Taylor, W.E.; Coles, G.A.; Shultz, M.V.; Egge, R.G.

    1993-09-01

    This paper describes an approach that facilitates management of personnel safety and environmental release risk from retired, surplus Westinghouse Hanford Company-managed facilities during the predemolition time frame. These facilities are located in the 100 and 200 Areas of the 1,450-km 2 (570-mi 2 ) Hanford Site in Richland, Washington. The production reactors are located in the 100 Area and the chemical separation facilities are located in the 200 Area. This paper also includes a description of the risk evaluation process, shows applicable results, and includes a description of comparison costs for different risk reduction options

  19. Westinghouse Hanford Company special nuclear material vault storage study

    International Nuclear Information System (INIS)

    Borisch, R.R.

    1996-01-01

    Category 1 and 2 Special Nuclear Materials (SNM) require storage in vault or vault type rooms as specified in DOE orders 5633.3A and 6430.1A. All category 1 and 2 SNM in dry storage on the Hanford site that is managed by Westinghouse Hanford Co (WHC) is located in the 200 West Area at Plutonium Finishing Plant (PFP) facilities. This document provides current and projected SNM vault inventories in terms of storage space filled and forecasts available space for possible future storage needs

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  1. Enhanced Westinghouse WWER-1000 fuel design for Ukraine reactors

    International Nuclear Information System (INIS)

    Dye, M.; Shah, H.

    2015-01-01

    Westinghouse has completed design, development, and region quantity delivery of an enhanced Westinghouse fuel assembly for WWER-1000 reactors to support continued safe reactor operations. The enhanced design builds on the successful performance of an earlier generation design which has operated in the South Ukraine 3 reactor for multiple cycles without any fuel rod failures. Incorporated design enhancements include a thicker spacer grid outer strap, an enhanced spacer grid outer strap profile to limit the risk for, and impact of, mechanical interaction/interference with coresident fuel, an all Alloy 718 grid structure for improved stability and strength, and improvements to the top and bottom nozzles. Capable of meeting increased lateral loads generated from using a higher axial trip limit for the refueling machine crane, the design was verified by extensive mechanical and thermalhydraulic testing, which included a newly developed fuel assembly-to-fuel assembly handling test rig to assess performance during bounding core loading and unloading conditions. Through these extensive design enhancements and comprehensive testing program, the enhanced WWER-1000 design provides additional performance, handling, and reliability margins for safe reactor operation. (authors)

  2. Westinghouse loading pattern search methodology for complex core designs

    International Nuclear Information System (INIS)

    Chao, Y.A.; Alsop, B.H.; Johansen, B.J.; Morita, T.

    1991-01-01

    Pressurized water reactor core designs have become more complex and must meet a plethora of design constraints. Trends have been toward longer cycles with increased discharge burnup, increased burnable absorber (BA) number, mixed BA types, reduced radial leakage, axially blanketed fuel, and multiple-batch feed fuel regions. Obtaining economical reload core loading patterns (LPs) that meet design criteria is a difficult task to do manually. Automated LP search tools are needed. An LP search tool cannot possibly perform an exhaustive search because of the sheer size of the combinatorial problem. On the other hand, evolving complexity of the design features and constraints often invalidates expert rules based on past design experiences. Westinghouse has developed a sophisticated loading pattern search methodology. This methodology is embodied in the LPOP code, which Westinghouse nuclear designers use extensively. The LPOP code generates a variety of LPs meeting design constraints and performs a two-cycle economic evaluation of the generated LPs. The designer selects the most appropriate patterns for fine tuning and evaluation by the design codes. This paper describes the major features of the LPOP methodology that are relevant to fulfilling the aforementioned requirements. Data and examples are also provided to demonstrate the performance of LPOP in meeting the complex design needs

  3. Disposition of weapons-grade plutonium in Westinghouse reactors

    International Nuclear Information System (INIS)

    Alsaed, A.A.; Adams, M.

    1998-03-01

    The authors have studied the feasibility of using weapons-grade plutonium in the form of mixed-oxide (MOX) fuel in existing Westinghouse reactors. They have designed three transition Cycles from an all LEU core to a partial MOX core. They found that four-loop Westinghouse reactors such as the Vogtle power plant are capable of handling up to 45 percent weapons-grade MOX loading without any modifications. The authors have also designed two kinds of weapons-grade MOX assemblies with three enrichments per assembly and four total enrichments. Wet annular burnable absorber (WABA) rods were used in all the MOX feed assemblies, some burned MOX assemblies, and some LEU feed assemblies. Integral fuel burnable absorber (IFBA) was used in the rest of the LEU feed assemblies. The average discharge burnup of MOX assemblies was over 47,000 MWD/MTM, which is more than enough to meet the open-quotes spent fuel standard.close quotes One unit is capable of consuming 0.462 MT of weapons-grade plutonium per year. Preliminary analyses showed that important reactor physics parameters for the three transitions cycles are comparable to those of LEU cores including boron levels, reactivity coefficients, peaking factors, and shutdown margins. Further transient analyses will need to be performed

  4. Validation of COMMIX with Westinghouse AP-600 PCCS test data

    International Nuclear Information System (INIS)

    Sun, J.G.; Chien, T.H.; Ding, J.; Sha, W.T.

    1993-01-01

    Small-scale test data for the Westinghouse AP-600 Passive Containment Cooling System (PCCS) have been used to validate the COMMIX computer code. To evaluate the performance of the PCCS, two transient liquid-film tracking models have been developed and implemented in the CO code. A set of heat transfer models and a mass transfer model based on heat and mass transfer analogy were used for the analysis of the AP-600 PCCS. It was found that the flow of the air stream in the annulus is a highly turbulent forced convection and that the flow of the air/steam mixture in the containment vessel is a mixed convection. Accordingly, a turbulent-forced-convection heat transfer model is used on the outside of the steel containment vessel wall and a mixed-convection heat transfer model is used on the inside of the steel containment vessel wall. The results from the CO calculations are compared with the experimental data from Westinghouse PCCS small-scale tests for average wall heat flux, evaporation rate, containment vessel pressure, and vessel wall temperature and heat flux distributions; agreement is good. The CO calculations also provide detailed distributions of velocity, temperature, and steam and air concentrations

  5. Performance of the Westinghouse WWER-1000 fuel design

    International Nuclear Information System (INIS)

    Höglund, J.; Jansson, A.; Latorre, R.; Davis, D.

    2015-01-01

    In 2005, six (6) Westinghouse WWER-1000 Lead Test Assemblies (LTAs) were loaded in South Ukraine Unit 3 (SU3). The LTAs completed the planned four cycles of operation and reached an average assembly burnup in excess of 43 MWd/ kgU. Post Irradiation Examination (PIE) inspections were performed after completion of each cycle and it was concluded that the 6 Westinghouse LTAs performed as expected during their operational regimes. In 2010, a full region of 42 assemblies of an enhanced WWER-1000 fuel design for Ukrainian reactors, designated WFA, was loaded in SU3. The WFA includes features that further mitigate assembly bow while at the same time improving the fuel cycle economy. In 2015, 26 WFAs completed their planned four cycles of operation reaching an average assembly burnup in excess of 42 MWd/ kgU. Currently 36 WFAs continue operating their fourth cycle in SU3. In addition, South Ukraine Unit 2 (SU2) has been loaded with WFAs and 27 assemblies have completed two cycles of operation reaching an average assembly burnup above 24 MWd/kgU. PIE for the WFAs has been completed after each cycle of operation. All assemblies have been examined for visible damage or non-standard position of fuel assembly components during unloading and reloading. All WFAs have also been subject to the standard leak testing process, with all fuel rods found to be hermetically sealed and non-leaking. Each outage, six WFAs have been subject to a more extensive inspection program. In 2012, 2013, and 2015, the Westinghouse Fuel Inspection and Repair Equipment (FIRE) workstation were used for the SU3 inspections. Excellent irradiation fuel performance has been observed and measured on all WFAs. The fuel assembly growth, rod cluster control assembly (RCCA) drag forces, oxide thickness, total fuel rod-to-nozzle gap channel closure, and fuel assembly bow data were within the bounds of the Westinghouse experience database. Results and concluding remarks from the PIEs are provided in this paper. In

  6. Westinghouse containment filtered venting system wet scrubber technology

    International Nuclear Information System (INIS)

    Kristensson, S.; Nilsson, P-O.

    2014-01-01

    Following the Fukushima event Westinghouse has further developed and enhanced its filtered containment venting system (FCVS) product line. The filtration efficiency of the proven FILTRA-MVSS system installed at all Swedish NPPs as well as at the Muhelberg plant in Switzerland has been enhanced and a new wet scrubber design, SVEN (Safety Venting), based on the FILTRA-MVSS tradition, developed. To meet increased filtration requirements for organic iodine these two wet scrubber products have been complemented with a zeolite module. The offering of a select choice of products allows for a better adjustment to the specific constraints and needs of each nuclear power station that is planning for the installation of such a system. The FILTRA-MVSS (MVSS=Multi Venturi Scrubber System) is a wet containment filtered vent system that uses multiple venturies to create an interaction between the vent gases and the scrubber media allowing for removal of aerosols and gaseous iodines in a very efficient manner. The FILTRA-MVSS was originally developed to meet stringent requirements on autonomy and maintained filtration efficiency over a wide range of venting conditions. The system was jointly developed in the late 80's by ABB Atom and ABB Flaekt, today Westinghouse and Alstom. Following installations in Sweden and Switzerland the system was further developed by replacement of the gravel-bed moisture separator with a standard demister and by addition of a set of sintered metal fibre filter cartridges placed after the moisture separator step. The system is today offered as a modular steel tank design to simplify installation at site. To reduce complexity and delivery time Westinghouse has developed an alternative design in which the venturi module is replaced by a submerged metal fibre filter cartridges module. This new wet scrubber design, SVEN (patent pending), provides a flexible, compact, and lower weight system, while still preserving and even enhancing the filtration

  7. The Westinghouse Waste Isolation Division Management and Supervisor Training Program

    International Nuclear Information System (INIS)

    Gilbreath, B.

    1992-01-01

    The Westinghouse Waste Isolation Division (WID) is the management and operating contractor (MOC) for the Department of Energy's (DOE's) Waste Isolation Plant (WIPP). Managers and supervisors at DOE facilities such as the WIPP are required to complete extensive training. To meet this requirement, WID created a self-paced, self-study program known as Management and Supervisor Training (MAST). All WID managers and supervisors are required to earn certification through the MAST program. Selected employees are permitted to participate in MAST with prior approval from their manager and the Human Resources Manager. Initial MAST certification requires the completion of 31 modules. MAST participants check out modules and read them when convenient. When they are prepared, participants take module examinations. To receive credit for a given module, participants must score at least 80 percent on the examination. Lessons learned from the development, implementation, and administration are presented in this paper

  8. Standard technical specifications for Westinghouse pressurized water reactors

    International Nuclear Information System (INIS)

    Wagner, P.C.

    1979-07-01

    This Standard Technical Specification (STS) has been structured for the broadest possible use on Westinghouse plants currently being reviewed for an Operating License. Accordingly, the document contains specifications applicable to plants with (1) either 3 or 4 loops and (2) with and without loop stop valves. In addition, four separate and discrete containment specification sections are provided for each of the following containment types: Atmospheric, Ice Condenser, Sub-Atmospheric, and Dual. Optional specifications are provided for those features and systems which may be included in individual plant designs but are not generic in their scope of application. Alternate specifications are provided in a limited number of cases to cover situations where alternate specification requirements are necessary on a generic basis because of design differences. This revision of the STS does not typically include requirements which may be added or revised as a result of the NRC staff's further review of the Three Mile Island incident

  9. Westinghouse Nuclear Core Design Training Center - a design simulator

    International Nuclear Information System (INIS)

    Altomare, S.; Pritchett, J.; Altman, D.

    1992-01-01

    The emergence of more powerful computing technology enables nuclear design calculations to be done on workstations. This shift to workstation usage has already had a profound effect in the training area. In 1991, the Westinghouse Electric Corporation's Commercial Nuclear Fuel Division (CNFD) developed and implemented a Nuclear Core Design Training Center (CDTC), a new concept in on-the-job training. The CDTC provides controlled on-the-job training in a structured classroom environment. It alllows one trainer, with the use of a specially prepared training facility, to provide full-scope, hands-on training to many trainees at one time. Also, the CDTC system reduces the overall cycle time required to complete the total training experience while also providing the flexibility of individual training in selected modules of interest. This paper provides descriptions of the CDTC and the respective experience gained in the application of this new concept

  10. The emergency response guidelines for the Westinghouse pressurized water reactor

    International Nuclear Information System (INIS)

    Dekens, J.P.; Bastien, R.; Prokopovich, S.R.

    1985-01-01

    The Three Mile Island accident has demonstrated that the guidance provided for mitigating the consequences of design basis accidents could be inadequate when multiple incidents, failures or errors occur during or after the accident. Westinghouse and the Westinghouse Owners Group have developed new Emergency Response Guidelines (E.R.G.). The E.R.G. are composed of two independent sets of procedures and of a systematic tool to continuously evaluate the plant safety throughout the response to an accident. a) The Optimal Recovery Guidelines are entered each time the reactor is tripped or the Emergency Core Cooling System is actuated. An immediate verification of the automatic protective actuations is performed and the accident diagnosis process is initiated. When nature of the accident is identified, the operator is transferred to the applicable recovery procedure and subprocedures. A permanent rediagnosis is performed throughout the application of the optimal Recovery Guidelines and cross connections are provided to the adequate procedure if an error in diagnosis is identified. b) Early in the course of the accident, the operating staff initiates monitoring of the Critical Safety Functions. These are defined as the set of functions ensuring the integrity of the physical barriers against radioactivity release. The review of these functions is peformed continuously through a cyclic application of the status trees. c) The Function Restoration Guidelines are entered when the Critical Safety Function monitoring identifies a challenge to one of the functions. Depending on the severity of the challenge, the transfer to a Function Restoration Guideline can be immediate for a severe challenge or delayed for a minor challenge. Those guidelines are independent of the scenario of the accident, but only based on plant parameters and equipment availability

  11. Application of realistic (best- estimate) methodologies for large break loss of coolant (LOCA) safety analysis: licensing of Westinghouse ASTRUM evaluation model in Spain

    International Nuclear Information System (INIS)

    Lage, Carlos; Frepoli, Cesare

    2010-01-01

    - and four loop Pressurized Water Reactor as well as passive plants (AP1000). This method follows the steps in the CSAU methodology. The uncertainty analysis (Element 3 in the CSAU) follows a direct Monte Carlo sampling procedure combined with non-parametric order statistics procedure. ASTRUM methodology is based on the use of WCOBRA/TRAC computer code which was validated against a large set of Separate and Integral Effects Tests (SETs and IETs). The objective of this paper is to discuss the main features of the ASTRUM Evaluation Model and its compliance with the Spanish regulatory requirements. (authors)

  12. Comparison of DNBR estimation methods in the Westinghouse and KWU reactor cores

    International Nuclear Information System (INIS)

    Camargo, C.T.M.; Pontedeiro, A.C.

    1984-11-01

    A method for foreseeing departure from nucleate boiling phenomenon in Westinghouse reator cores (OTΔT- signal for reator shut down) is described. The results from investigations done with the OTΔT system and in the efficiency of different methods used in the Westinghouse and KWU nuclear power plants to estimate thermohydraulic conditions of the PWR reactor cores, are presented. The investigations were done, by support of computer codes. The modifications, purposed by Westinghouse, in the original project of Angra-1 OTΔT system are analysed. (M.C.K.) [pt

  13. Review of Reliability Assessment of Westinghouse SSPS Using SPC by WEC

    International Nuclear Information System (INIS)

    Kang, H. T.; Chung, H. Y.

    2007-01-01

    Westinghouse Electric Company (WEC) has accomplished the reliability assessment of Westinghouse Solid State Protection System (SSPS) in KORI no. 2, 3, 4, and YGN no. 1, 2. In their studies, it is reported that creating a cost-effective plan for improving the reliability of the SSPS and at KORI no. 2, 3 and 4, and YGN no. 1, 2 should be needed while reducing their maintenance cost. In this paper, we reviewed the reliability assessment of Westinghouse SSPS analyzed in two performance standards, availability, and the maintenance expense using Statistic Process Control (SPC). As a result, it is concluded all plants have several failures reported but no effect on the system's availability, and the maintenance expense analysis did not reduce the current maintenance expense by 30%. Therefore, overall review for the reliability assessment is that a new strategy for cost-effective plan and/or upgrade approach for improving the reliability of the aging Westinghouse SSPS should be needed

  14. Validating Westinghouse atom 16 x 16 and 18 x 18 PWR fuel performance

    International Nuclear Information System (INIS)

    Andersson, S.; Gustafson, J.; Jourdain, P.; Lindstroem, L.; Hallstadius, L.; Hofling, C.G.

    2001-01-01

    Westinghouse Atom designs and fabricates PWR fuel for all major European fuel types: 17 x 17 standard (12 ft) and 17 x 17 XL (14 ft) for Westinghouse type PWRs, and 16 x 16 and 18 x 18 fuel for Siemens type PWRs. The W Atom PWR fuel designs are based on the extensive Westinghouse CE PWR fuel experience from combustion engineering type PWRs. The W atom designs utilise basic design features from the W CE fuel tradition, such as all-Zircaloy mid grids and the proven ( 6 rod years) Guardian TM debris catcher, which is integrated in the bottom Inconel grid. Several new features have been developed to meet with stringent European requirements originating from requirements on very high burnup, in combination with low-leakage core operating strategies and high coolant temperatures. The overall reliability of the Westinghouse Atom PWR fuel is very high; no fuel failure has been detected since 1997. (orig.)

  15. A Westinghouse designed distributed mircroprocessor based protection and control system

    International Nuclear Information System (INIS)

    Bruno, J.; Reid, J.B.

    1980-01-01

    For approximately five years, Westinghouse has been involved in the design and licensing of a distributed microprocessor based system for the protection and control of a pressurized water reactor nuclear steam supply system. A 'top-down' design methodology was used, in which the system global performance objectives were specified, followed by increasingly more detailed design specifications which ultimately decomposed the system into its basic hardware and software elements. The design process and design decisions were influenced by the recognition that the final product would have to be verified to ensure its capability to perform the safety-related functions of a class 1E protection system. The verification process mirrored the design process except that it was 'bottom-up' and thus started with the basic elements and worked upwards through the system in increasingly complex blocks. A number of areas which are of interest in a distributed system are disucssed, with emphasis on two systems. The first, the Integrated Protection System is primarily responsible for processing signals from field mounted sensors to provide for reactor trips and the initiation of the Engineered Safety Features. The Integrated Control System, which is organized in a parallel manner, processes other sensor signals and generates the necessary analog and on-off signals to maintain the plant parameters within specified limits. Points covered include system structure, systems partitioning strategies, communications techniques, software design concepts, reliability and maintainability, commercial component availability, interference susceptibility, licensing issues, and applicability. (LL)

  16. The Westinghouse BEACON on-line core monitoring system

    International Nuclear Information System (INIS)

    Buechel, Robert J.; Boyd, William A.; Casadei, Alberto L.

    1995-01-01

    BEACON (Best Estimate Analysis of Core Operations - Nuclear), a core monitoring and operational support package developed by Westinghouse, has been installed at many operating PWRs worldwide. The BEACON system is a real-time monitoring system which can be used in plants with both fixed and movable incore detector systems and utilizes an on-line nodal model combined with core instrumentation data to provide continuous core power distribution monitoring. In addition, accurate core-predictive capabilities utilizing a full core nodal model updated according to plant operating history can be made to provide operational support. Core history information is kept and displayed to help operators anticipate core behavior and take pro-active control actions. The BEACON system has been licensed by the U.S. Nuclear Regulatory Commission for direct, continuous monitoring of DNBR and peak linear heat rate. This allows BEACON to be integrated into the plant technical specifications to permit significant relaxation of operating limitations defined by conventional technical specifications. (author). 4 refs, 2 figs, 1 tab

  17. Westinghouse Hanford Company Operational Environmental Monitoring. Annual report, CY 1993

    International Nuclear Information System (INIS)

    Schmidt, J.W.; Johnson, A.R.; Markes, B.M.; McKinney, S.M.; Perkins, C.J.

    1994-07-01

    This document presents the results of the Westinghouse Hanford Company near-facility operational environmental monitoring for 1993 in the 100, 200/600, and 300/400 Areas of the Hanford Site, in south-central Washington State. Surveillance activities included sampling and analyses of ambient air, surface water, groundwater, sediments, soil, and biota. Also, external radiation measurements and radiological surveys were taken at waste disposal sites, radiologically controlled areas, and roads. These activities were conducted to assess and control the effects of nuclear facilities and waste sites on the local environment. In addition, diffuse sources were monitored to determine compliance with Federal, State, and/or local regulations. In general, although effects from nuclear facilities are still seen on the Hanford Site and radiation levels are slightly elevated when compared to offsite conditions, the differences are less than in previous years. At certain locations on or directly adjacent to nuclear facilities and waste sites, levels can be several times higher than offsite conditions

  18. A Study on Dismantling of Westinghouse Type Nuclear Reactor

    International Nuclear Information System (INIS)

    Jeong, Woo-Tae; Lee, Sang-Guk

    2014-01-01

    KHNP started a research project this year to develop a methodology to dismantle nuclear reactors and internals. In this paper, we reviewed 3D design model of the reactor and suggested feasible cutting scheme.. Using 3-D CAD model of Westinghouse type nuclear reactor and its internals, we reviewed possible options for disposal. Among various options of dismantling the nuclear reactor, plasma cutting was selected to be the best feasible and economical method. The upper internals could be segmented by using a band saw. It is relatively fast, and easily maintained. For cutting the lower internals, plasma torch was chosen to be the best efficient tool. Disassembling the baffle and the former plate by removing the baffle former bolts was also recommended for minimizing storage volume. When using plasma torch for cutting the reactor vessel and its internal, installation of a ventilation system for preventing pollution of atmosphere was recommended. For minimizing radiation exposure during the cutting operation, remotely controlled robotic tool was recommended to be used

  19. Westinghouse Savannah River Company (WSRC) approach to nuclear facility maintenance

    International Nuclear Information System (INIS)

    Harrison, D.W.

    1991-01-01

    The Savannah River Site (SRS) in South Carolina is a 300+ square mile facility owned by the US Department of Energy (DOE) and operated by Westinghouse Savannah River Company (WSRC), the prime contractor; Bechtel Savannah River, Incorporated (BSRI) is a major subcontractor. The site has used all of the five nuclear reactors and it has the necessary nuclear materials processing facilities, as well as waste management and research facilities. The site has produced materials for the US nuclear arsenal and various isotopes for use in space research and nuclear medicine for more than 30 years. In 1989, WSRC took over as prime contractor, replacing E.I. du Pont de Nemours and Company. At this time, a concentrated effort began to more closely align the operating standards of this site with those accepted by the commercial nuclear industry of the United States. Generally, this meant acceptance of standards of the Institute of Nuclear Power Operations (INPO) for nuclear-related facilities at the site. The subject of this paper is maintenance of nuclear facilities and, therefore, excludes discussion of the maintenance of non-nuclear facilities and equipment

  20. Standard technical specifications for Westinghouse pressurized water reactors (revision issued Fall 1981). Technical report

    International Nuclear Information System (INIS)

    Virgilio, M.J.

    1981-11-01

    The Standard Technical Specifications for Westinghouse Pressurized Water Reactors (W-STS) is a generic document prepared by the U.S. NRC for use in the licensing process of current Westinghouse Pressurized Water Reactors. The W-STS sets forth the Limits, Operating Conditions and other requirements applicable to nuclear reactor facility operation as set forth in Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public

  1. Age-related degradation of Westinghouse 480-volt circuit breakers

    International Nuclear Information System (INIS)

    Subudhi, M.; MacDougall, E.; Kochis, S.; Wilhelm, W.; Lee, B.S.

    1990-11-01

    After the McGuire event in 1987 relating to failure of the center pole weld in one of its reactor trip breakers, activities were initiated by the NRC to investigate the probable causes. A review of operating experience suggested that the burning of coils, jamming of the operating mechanism, and deterioration of the contacts dominated the breakers failures. Although failures of the pole shaft weld were not included as one of the generic problems, the NRC augmented inspection team had suspected that these welds were substandard which led them to crack prematurely. A DS-416 low voltage air circuit breaker manufactured by Westinghouse was mechanically cycled to identify age-related degradations. This accelerated aging test was conducted for over 36,000 cycles during nine months. Three separate pole shafts, one with a 60 degree weld, one with a 120 degree and one with a 180 degree were used to characterize the cracking in the pole level welds. In addition, three different operating mechanisms and several other parts were replaced as they became inoperable. The testing yielded many useful results. The burning of the closing coils was found to be the effect of binding in the linkages that are connected to this device. Among the seven welds on the pole shaft, number-sign 1 and number-sign 3 were the critical ones which cracked first to cause misalignment of the pole levers, which, in turn, had led to many problems with the operating mechanism including the burning of coils, excessive wear in certain parts, and overstressed linkages. Based on these findings, a maintenance program is suggested to alleviate the age-related degradations that occur due to mechanical cycling of this type of breaker. 3 refs., 39 figs., 7 tabs

  2. Westinghouse modular grinding process - improvement for follow on processes

    International Nuclear Information System (INIS)

    Fehrmann, Henning

    2013-01-01

    In nuclear power plants (NPP) ion exchange (IX) resins are used in several systems for water treatment. The resins can be in bead or powdered form. For waste treatment of spent IX resins, two methods are basically used: Direct immobilization (e.g. with cement, bitumen, polymer or High Integrity Container (HIC)); Thermal treatment (e.g. drying, oxidation or pyrolysis). Bead resins have some properties (e.g. particle size and density) that can have negative impacts on following waste treatment processes. Negative impacts could be: Floatation of bead resins in cementation process; Sedimentation in pipeline during transportation; Poor compaction properties for Hot Resin Supercompaction (HRSC). Reducing the particle size of the bead resins can have beneficial effects enhancing further treatment processes and overcoming prior mentioned effects. Westinghouse Electric Company has developed a modular grinding process to crush/grind the bead resins. This modular process is designed for flexible use and enables a selective adjustment of particle size to tailor the grinding system to the customer needs. The system can be equipped with a crusher integrated in the process tank and if necessary a colloid mill. The crusher reduces the bead resins particle size and converts the bead resins to a pump able suspension with lower sedimentation properties. With the colloid mill the resins can be ground to a powder. Compared to existing grinding systems this equipment is designed to minimize radiation exposure of the worker during operation and maintenance. Using the crushed and/or ground bead resins has several beneficial effects like facilitating cementation process and recipe development, enhancing oxidation of resins, improving the Hot Resin Supercompaction volume reduction performance. (authors)

  3. Westinghouse modular grinding process - improvement for follow on processes

    Energy Technology Data Exchange (ETDEWEB)

    Fehrmann, Henning [Westinghouse Germany GmbH, Mannheim, State (Germany)

    2013-07-01

    In nuclear power plants (NPP) ion exchange (IX) resins are used in several systems for water treatment. The resins can be in bead or powdered form. For waste treatment of spent IX resins, two methods are basically used: Direct immobilization (e.g. with cement, bitumen, polymer or High Integrity Container (HIC)); Thermal treatment (e.g. drying, oxidation or pyrolysis). Bead resins have some properties (e.g. particle size and density) that can have negative impacts on following waste treatment processes. Negative impacts could be: Floatation of bead resins in cementation process; Sedimentation in pipeline during transportation; Poor compaction properties for Hot Resin Supercompaction (HRSC). Reducing the particle size of the bead resins can have beneficial effects enhancing further treatment processes and overcoming prior mentioned effects. Westinghouse Electric Company has developed a modular grinding process to crush/grind the bead resins. This modular process is designed for flexible use and enables a selective adjustment of particle size to tailor the grinding system to the customer needs. The system can be equipped with a crusher integrated in the process tank and if necessary a colloid mill. The crusher reduces the bead resins particle size and converts the bead resins to a pump able suspension with lower sedimentation properties. With the colloid mill the resins can be ground to a powder. Compared to existing grinding systems this equipment is designed to minimize radiation exposure of the worker during operation and maintenance. Using the crushed and/or ground bead resins has several beneficial effects like facilitating cementation process and recipe development, enhancing oxidation of resins, improving the Hot Resin Supercompaction volume reduction performance. (authors)

  4. Westinghouse Hanford Company Pollution Prevention Program Implementation Plan

    International Nuclear Information System (INIS)

    Floyd, B.C.

    1994-10-01

    This plan documents Westinghouse Hanford Company's (WHC) Pollution Prevention (P2) (formerly Waste Minimization) program. The program includes WHC; BCS Richland, Inc. (BCSR); and ICF Kaiser Hanford Company (ICF KH). The plan specifies P2 program activities and schedules for implementing the Hanford Site Waste Minimization and Pollution Prevention Awareness (WMin/P2) Program Plan requirements (DOE 1994a). It is intended to satisfy the U.S. Department of Energy (DOE) and other legal requirements that are discussed in both the Hanford Site WMin/P2 plan and paragraph C of this plan. As such, the Pollution Prevention Awareness Program required by DOE Order 5400.1 (DOE 1988) is included in the WHC P2 program. WHC, BCSR, and ICF KH are committed to implementing an effective P2 program as identified in the Hanford Site WMin/P2 Plan. This plan provides specific information on how the WHC P2 program will develop and implement the goals, activities, and budget needed to accomplish this. The emphasis has been to provide detailed planning of the WHC P2 program activities over the next 3 years. The plan will guide the development and implementation of the program. The plan also provides background information on past program activities. Because the plan contains greater detail than in the past, activity scope and implementation schedules may change as new priorities are identified and new approaches are developed and realized. Some activities will be accelerated, others may be delayed; however, all of the general program elements identified in this plan and contractor requirements identified in the Site WMin/P2 plan will be developed and implemented during the next 3 years. This plan applies to all WHC, BCSR, and ICF KH organizations and subcontractors. It will be distributed to those with defined responsibilities in this plan; and the policy, goals, objectives, and strategy of the program will be communicated to all WHC, BCSR, and ICF KH employees

  5. Age-related degradation of Westinghouse 480-volt circuit breakers

    International Nuclear Information System (INIS)

    Subudhi, M.; Shier, W.; MacDougall, E.

    1990-07-01

    An aging assessment of Westinghouse DS-series low-voltage air circuit breakers was performed as part of the Nuclear Plant Aging Research (NPAR) program. The objectives of this study are to characterize age-related degradation within the breaker assembly and to identify maintenance practices to mitigate their effect. Since this study has been promulgated by the failures of the reactor trip breakers at the McGuire Nuclear Station in July 1987, results relating to the welds in the breaker pole lever welds are also discussed. The design and operation of DS-206 and DS-416 breakers were reviewed. Failure data from various national data bases were analyzed to identify the predominant failure modes, causes, and mechanisms. Additional operating experiences from one nuclear station and two industrial breaker-service companies were obtained to develop aging trends of various subcomponents. The responses of the utilities to the NRC Bulletin 88-01, which discusses the center pole lever welds, were analyzed to assess the final resolution of failures of welds in the reactor trips. Maintenance recommendations, made by the manufacturer to mitigate age-related degradation were reviewed, and recommendations for improving the monitoring of age-related degradation are discussed. As described in Volume 2 of this NUREG, the results from a test program to assess degradation in breaker parts through mechanical cycling are also included. The testing has characterized the cracking of center-pole lever welds, identified monitoring techniques to determine aging in breakers, and provided information to augment existing maintenance programs. Recommendations to improve breaker reliability using effective maintenance, testing, and inspection programs are suggested. 13 refs., 21 figs., 8 tabs

  6. Non-chemical water purification a Westinghouse/Wallenius product for nuclear power plant needs

    International Nuclear Information System (INIS)

    Goetberg, J.; Carlsson, M.

    2014-01-01

    Increasing demand for ecologically effective water treatment technologies has resulted in the development of several new oxidation methods. These technologies are generally labelled Advanced Oxidation Technologies (AOT) or Advanced Oxidation Processes (AOP) and currently represent the most widely recognized alternative for ecologically sound, high-tech water purification. Many years of intensive research have culminated in the innovative Wallenius-AOT technology, a patented method that is remarkable in several ways. It imitates nature's own water purification method. This means no chemical additives are needed. The technology utilizes the ability of light, together with photo-catalytic semiconductor surfaces, to produce free radicals, like nature does. These reactive radicals create an environment in which organic and inorganic substances oxidize, whereby a broad spectrum of organisms is rendered harmless more effectively than with conventional UV technology. The entire process takes just a few micro-seconds. A major advantage of the technology is that it can be adjusted according to the desired degree of purification. By altering the dynamics of the process, the purification can be designed for specific applications. In this way, AOT tackles precise problems, regardless of flow and whether the problem is chemical or biological. The product was originally introduced for ballast treatment in the shipping industry. Ballast water has created severe damages to the biology at many locations. By moving an organism from one ocean to another we have introduced a possible threat to the local ecosystem. This has been prevented by using the AOT water treatment units. During ballasting and de-ballasting, the units create radicals with the help of a catalyst and a light source. These radicals then destroy the cell membrane of microorganisms. The radicals, which never leave the unit, have a lifetime of only a few milliseconds and pose no risk to the environment or crew

  7. Westinghouse Small Modular Reactor balance of plant and supporting systems design

    Energy Technology Data Exchange (ETDEWEB)

    Memmott, M. J.; Stansbury, C.; Taylor, C. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the second in a series of four papers which describe the design and functionality of the Westinghouse SMR. It focuses, in particular, upon the supporting systems and the balance of plant (BOP) designs of the Westinghouse SMR. Several Westinghouse SMR systems are classified as safety, and are critical to the safe operation of the Westinghouse SMR. These include the protection and monitoring system (PMS), the passive core cooling system (PXS), and the spent fuel cooling system (SFS) including pools, valves, and piping. The Westinghouse SMR safety related systems include the instrumentation and controls (I and C) as well as redundant and physically separated safety trains with batteries, electrical systems, and switch gears. Several other incorporated systems are non-safety related, but provide functions for plant operations including defense-in-depth functions. These include the chemical volume control system (CVS), heating, ventilation and cooling (HVAC) systems, component cooling water system (CCS), normal residual heat removal system (RNS) and service water system (SWS). The integrated performance of the safety-related and non-safety related systems ensures the safe and efficient operation of the Westinghouse SMR through various conditions and transients. The turbine island consists of the turbine, electric generator, feedwater and steam systems, moisture separation systems, and the condensers. The BOP is designed to minimize assembly time, shipping challenges, and on-site testing requirements for all structures, systems, and components. (authors)

  8. Westinghouse Small Modular Reactor balance of plant and supporting systems design

    International Nuclear Information System (INIS)

    Memmott, M. J.; Stansbury, C.; Taylor, C.

    2012-01-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the second in a series of four papers which describe the design and functionality of the Westinghouse SMR. It focuses, in particular, upon the supporting systems and the balance of plant (BOP) designs of the Westinghouse SMR. Several Westinghouse SMR systems are classified as safety, and are critical to the safe operation of the Westinghouse SMR. These include the protection and monitoring system (PMS), the passive core cooling system (PXS), and the spent fuel cooling system (SFS) including pools, valves, and piping. The Westinghouse SMR safety related systems include the instrumentation and controls (I and C) as well as redundant and physically separated safety trains with batteries, electrical systems, and switch gears. Several other incorporated systems are non-safety related, but provide functions for plant operations including defense-in-depth functions. These include the chemical volume control system (CVS), heating, ventilation and cooling (HVAC) systems, component cooling water system (CCS), normal residual heat removal system (RNS) and service water system (SWS). The integrated performance of the safety-related and non-safety related systems ensures the safe and efficient operation of the Westinghouse SMR through various conditions and transients. The turbine island consists of the turbine, electric generator, feedwater and steam systems, moisture separation systems, and the condensers. The BOP is designed to minimize assembly time, shipping challenges, and on-site testing requirements for all structures, systems, and components. (authors)

  9. Regulatory analysis for the resolution of Generic Issue 115, enhancement of the reliability of the Westinghouse Solid State Protection System

    International Nuclear Information System (INIS)

    Basdekas, D.L.

    1989-05-01

    Generic Issue 115 addresses a concern related to the reliability of the Westinghouse reactor protection system for plants using the Westinghouse Solid State Protection System (SSPS). Several options for improving the reliability of the Westinghouse reactor trip function for these plants and their effect on core damage frequency (CDF) and overall risk were evaluated. This regulatory analysis includes a quantitative assessment of the costs and benefits associated with the various options for enhancing the reliability of the Westinghouse SSPS and provides insights for consideration and industry initiatives. No new regulatory requirements are proposed. 25 refs., 11 tabs

  10. Westinghouse Hanford Company waste minimization and pollution prevention awareness program plan

    International Nuclear Information System (INIS)

    Craig, P.A.; Nichols, D.H.; Lindsey, D.W.

    1991-08-01

    The purpose of this plan is to establish the Westinghouse Hanford Company's Waste Minimization Program. The plan specifies activities and methods that will be employed to reduce the quantity and toxicity of waste generated at Westinghouse Hanford Company (Westinghouse Hanford). It is designed to satisfy the US Department of Energy (DOE) and other legal requirements that are discussed in Subsection C of the section. The Pollution Prevention Awareness Program is included with the Waste Minimization Program as permitted by DOE Order 5400.1 (DOE 1988a). This plan is based on the Hanford Site Waste Minimization and Pollution Prevention Awareness Program Plan, which directs DOE Field Office, Richland contractors to develop and maintain a waste minimization program. This waste minimization program is an organized, comprehensive, and continual effort to systematically reduce waste generation. The Westinghouse Hanford Waste Minimization Program is designed to prevent or minimize pollutant releases to all environmental media from all aspects of Westinghouse Hanford operations and offers increased protection of public health and the environment. 14 refs., 2 figs., 1 tab

  11. WEOD-S: Westinghouse expanded operating domain stability solution

    International Nuclear Information System (INIS)

    Rotander, C.; Blaisdell, J.; Anderson, D.; Kumar, V.; Stier, D.; Chu, E.

    2014-01-01

    As Extended Power up-rates (EPUs) are implemented in BWR plants, the flow window at full power decreases due to the extension of the rod line. It is thus desirable to raise load line limits to realize increased power generation at a wider flow range offering operational flexibility and fuel cycle efficiency. However, when load lines are raised, the power/flow operating map is changed in a direction that can cause core power instability at its lower left corner (high power/low flow) if a flow reduction transient (i.e. pump trip) occurs. Unstable operation of the reactor core can result in diverging neutron flux (and power) oscillations, and through the thermal hydraulic/neutronic feedback challenge the Safety Limit Minimum Critical Power Ratio (SLMCPR). In many BWRs the SLMCPR in a power oscillation event is already protected by a detect and suppress system. The methodology to determine the set point of this system, the DIVOM methodology (Delta CPR over Initial MCPR versus Oscillation Magnitude), is defined and applicable up to, but not beyond, the thermal hydraulic stability limit. The DIVOM methodology is used to determine the channel power oscillation magnitude that will challenge the SLMCPR. It is defined as the relationship between ΔCPR/ICPR and the Hot Channel Oscillation Magnitude (HCOM). The DIVOM calculations are typically performed at the end state following a design basis two pump trip from rated power and minimum flow. When approaching the thermal hydraulic (T/H) instability limit, the DIVOM curve can become chaotic and the DIVOM approach breaks down. At T/H-instability, small power fluctuations give rise to large flow oscillations and the non-linear dynamic properties emerge. The newly developed Westinghouse Expanded Operating Domain Stability (WEOD-S) solution proactively prevents entry into the regions of the power/flow map that are vulnerable to thermal hydraulic instability. This is achieved automatically, without any dependence on operator action

  12. Westinghouse corporate development of a decision software program for Radiological Evaluation Decision Input (REDI)

    International Nuclear Information System (INIS)

    Bush, T.S.

    1995-01-01

    In December 1992, the Department of Energy (DOE) implemented the DOE Radiological Control Manual (RCM). Westinghouse Idaho Nuclear Company, Inc. (WINCO) submitted an implementation plan showing how compliance with the manual would be achieved. This implementation plan was approved by DOE in November 1992. Although WINCO had already been working under a similar Westinghouse RCM, the DOE RCM brought some new and challenging requirements. One such requirement was that of having procedure writers and job planners create the radiological input in work control procedures. Until this time, that information was being provided by radiological engineering or a radiation safety representative. As a result of this requirement, Westinghouse developed the Radiological Evaluation Decision Input (REDI) program

  13. Westinghouse corporate development of a decision software program for Radiological Evaluation Decision Input (REDI)

    Energy Technology Data Exchange (ETDEWEB)

    Bush, T.S. [Westinghosue Idaho Nuclear Co., Inc., Idaho Falls, ID (United States)

    1995-03-01

    In December 1992, the Department of Energy (DOE) implemented the DOE Radiological Control Manual (RCM). Westinghouse Idaho Nuclear Company, Inc. (WINCO) submitted an implementation plan showing how compliance with the manual would be achieved. This implementation plan was approved by DOE in November 1992. Although WINCO had already been working under a similar Westinghouse RCM, the DOE RCM brought some new and challenging requirements. One such requirement was that of having procedure writers and job planners create the radiological input in work control procedures. Until this time, that information was being provided by radiological engineering or a radiation safety representative. As a result of this requirement, Westinghouse developed the Radiological Evaluation Decision Input (REDI) program.

  14. Feedback from Westinghouse experience on segmentation of reactor vessel internals - 59013

    International Nuclear Information System (INIS)

    Kreitman, Paul J.; Boucau, Joseph; Segerud, Per; Fallstroem, Stefan

    2012-01-01

    With more than 25 years of experience in the development of reactor vessel internals segmentation and packaging technology, Westinghouse has accumulated significant know-how in the reactor dismantling market. Building on tooling concepts and cutting methodologies developed decades ago for the successful removal of nuclear fuel from the damaged Three Mile Island Unit 2 reactor (TMI-2), Westinghouse has continuously improved its approach to internals segmentation and packaging by incorporating lessons learned and best practices into each successive project. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive water-jet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Westinghouse has applied its technology to all types of reactors covering Pressurized Water Reactors (PWR's), Boiling Water Reactors (BWR's), Gas Cooled Reactors (GCR's) and sodium reactors. The primary challenges of a segmentation and packaging project are to separate the highly activated materials from the less-activated materials and package them into appropriate containers for disposal. Since space is almost always a limiting factor it is therefore important to plan and optimize the available room in the segmentation areas. The choice of the optimum cutting technology is important for a successful project implementation and depends on some specific constraints like disposal costs, project schedule, available areas or safety. Detailed 3-D modeling is the basis for tooling design and provides invaluable support in determining the optimum strategy for component cutting and disposal in waste containers, taking account of the radiological and packaging constraints. Westinghouse has also developed a variety of special handling tools, support fixtures, service bridges, water filtration systems, video-monitoring systems and customized rigging, all of which are required for a

  15. Sensitivity Analysis on LOCCW of Westinghouse typed Reactors Considering WOG2000 RCP Seal Leakage Model

    International Nuclear Information System (INIS)

    Na, Jang-Hwan; Jeon, Ho-Jun; Hwang, Seok-Won

    2015-01-01

    In this paper, we focus on risk insights of Westinghouse typed reactors. We identified that Reactor Coolant Pump (RCP) seal integrity is the most important contributor to Core Damage Frequency (CDF). As we reflected the latest technical report; WCAP-15603(Rev. 1-A), 'WOG2000 RCP Seal Leakage Model for Westinghouse PWRs' instead of the old version, RCP seal integrity became more important to Westinghouse typed reactors. After Fukushima accidents, Korea Hydro and Nuclear Power (KHNP) decided to develop Low Power and Shutdown (LPSD) Probabilistic Safety Assessment (PSA) models and upgrade full power PSA models of all operating Nuclear Power Plants (NPPs). As for upgrading full power PSA models, we have tried to standardize the methodology of CCF (Common Cause Failure) and HRA (Human Reliability Analysis), which are the most influential factors to risk measures of NPPs. Also, we have reviewed and reflected the latest operating experiences, reliability data sources and technical methods to improve the quality of PSA models. KHNP has operating various types of reactors; Optimized Pressurized Reactor (OPR) 1000, CANDU, Framatome and Westinghouse. So, one of the most challengeable missions is to keep the balance of risk contributors of all types of reactors. This paper presents the method of new RCP seal leakage model and the sensitivity analysis results from applying the detailed method to PSA models of Westinghouse typed reference reactors. To perform the sensitivity analysis on LOCCW of the reference Westinghouse typed reactors, we reviewed WOG2000 RCP seal leakage model and developed the detailed event tree of LOCCW considering all scenarios of RCP seal failures. Also, we performed HRA based on the T/H analysis by using the leakage rates for each scenario. We could recognize that HRA was the sensitive contributor to CDF, and the RCP seal failure scenario of 182gpm leakage rate was estimated as the most important scenario

  16. Sensitivity Analysis on LOCCW of Westinghouse typed Reactors Considering WOG2000 RCP Seal Leakage Model

    Energy Technology Data Exchange (ETDEWEB)

    Na, Jang-Hwan; Jeon, Ho-Jun; Hwang, Seok-Won [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this paper, we focus on risk insights of Westinghouse typed reactors. We identified that Reactor Coolant Pump (RCP) seal integrity is the most important contributor to Core Damage Frequency (CDF). As we reflected the latest technical report; WCAP-15603(Rev. 1-A), 'WOG2000 RCP Seal Leakage Model for Westinghouse PWRs' instead of the old version, RCP seal integrity became more important to Westinghouse typed reactors. After Fukushima accidents, Korea Hydro and Nuclear Power (KHNP) decided to develop Low Power and Shutdown (LPSD) Probabilistic Safety Assessment (PSA) models and upgrade full power PSA models of all operating Nuclear Power Plants (NPPs). As for upgrading full power PSA models, we have tried to standardize the methodology of CCF (Common Cause Failure) and HRA (Human Reliability Analysis), which are the most influential factors to risk measures of NPPs. Also, we have reviewed and reflected the latest operating experiences, reliability data sources and technical methods to improve the quality of PSA models. KHNP has operating various types of reactors; Optimized Pressurized Reactor (OPR) 1000, CANDU, Framatome and Westinghouse. So, one of the most challengeable missions is to keep the balance of risk contributors of all types of reactors. This paper presents the method of new RCP seal leakage model and the sensitivity analysis results from applying the detailed method to PSA models of Westinghouse typed reference reactors. To perform the sensitivity analysis on LOCCW of the reference Westinghouse typed reactors, we reviewed WOG2000 RCP seal leakage model and developed the detailed event tree of LOCCW considering all scenarios of RCP seal failures. Also, we performed HRA based on the T/H analysis by using the leakage rates for each scenario. We could recognize that HRA was the sensitive contributor to CDF, and the RCP seal failure scenario of 182gpm leakage rate was estimated as the most important scenario.

  17. Westinghouse calls for rethink on Europe's treatment of nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Kraev, Kamen [NucNet The Independent Global Nuclear News Agency, Brussels (Belgium)

    2017-12-15

    US-based nuclear equipment manufacturer Westinghouse Electric Company has called on European Union legislators to adopt a technology-neutral approach when discussing the future of the bloc's low-carbon energy policies. In its 'Clean Energy for All Europeans' legislative package, released in November 2016, the European Commission made no mention of nuclear energy, said Michael Kirst, Westinghouse's vice-president of strategy for the Europe, Middle East and Africa (EMEA) region at a media briefing in Brussels. He said the package did not offer ''a real investment signal'' to developers.

  18. 1992 Environmental Summer Science Camp Program evaluation. The International Environmental Institute of Westinghouse Hanford Company

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    This report describes the 1992 Westinghouse Hanford Company/US Department of Energy Environmental Summer Science Camp. The objective of the ``camp`` was to motivate sixth and seventh graders to pursue studies in math, science, and the environment. This objective was accomplished through hands-on fun activities while studying the present and future challenges facing our environment. The camp was funded through Technical Task Plan, 424203, from the US Department of Energy-Headquarters, Office of Environmental Restoration and Waste Management, Technology Development,to Westinghouse Hanford Company`s International Environmental Institute, Education and Internship Performance Group.

  19. Information storage and retrieval system at Westinghouse Hanford Company Hanford Engineering Development Laboratory (HEDL)

    International Nuclear Information System (INIS)

    Theo, M.G.

    1977-01-01

    The information storage and retrieval system developed at Westinghouse--Hanford is described. It will be able to store over two million documents on line. The system uses an interactive minicomputer to search for keyworded documents. Documents of interest can be displayed on CRTs or printed on microfilm reader--printers. 31 figures

  20. Instructional skills training - the Westinghouse program to insure competence of nuclear training instructors

    International Nuclear Information System (INIS)

    Widen, W.C.

    1983-01-01

    The nuclear training engineer as well as being competent technically must be able to teach effectively. Westinghouse have developed a course for training instructors which aims to improve their teaching skills. The course, which has both theoretical and practical content covers the role of the instructor, the learning process, communications, test construction and analysis and stress identification and analysis. (U.K.)

  1. Westinghouse Hanford Company (WHC) standards/requirements identification document (S/RID)

    International Nuclear Information System (INIS)

    Bennett, G.L.

    1996-01-01

    This Standards/Requirements Identification Document (S/RID) set forth the Environmental Safety and Health (ES ampersand amp;H) standards/requirements for Westinghouse Hanford Company Level Programs, where implementation and compliance is the responsibility of these organizations. These standards/requirements are adequate to ensure the protection of the health and safety of workers, the public, and the environment

  2. The Westinghouse approach - an I and C modernization program for WWERs

    International Nuclear Information System (INIS)

    Werner, C.L.; Wassel, W.W.; Novak, V.

    1993-01-01

    When entering into a design program that is a marriage between two designs it is very difficult to separate self imposed design criteria from the requirements of the program. Therefore, the criteria of and the requirements for the Westinghouse modernization program will be discussed as one. These are outlined below: 1) The OSART Mission that was conducted by the IAEA at the Temelin Plant in 1990 identified the need to provide a new comprehensive Safety Analysis to verify the various aspects of the WWER safety system design. This recommendation is one that Westinghouse will provide as part of the WWER I and C Modernization Program. The design, no matter how well proven or verified from a hardware design point of view, is only as good as the basis for the system design; 2) Minimize the impact on the civil design aspects of the plant where possible and where this requirements do not affect the safety features of the design; 3) Ensure compatibility of the design to meet the latest US NRC requirements and those of the implementing country, applicable to the systems functional and hardware designs. This is a Westinghouse standard corporate requirement for all nuclear plant and systems design whether they be foreign or domestic; 4) Provide the most modern, proven design for the I and C systems. Application of the Westinghouse Instrumentation and Control microprocessor based design to the WWER Modernization Program will provide the basis for upgrading plants to meet western standards. (author) 6 figs., 1 ref

  3. WIMSD4 calculations of the Westinghouse 'EDASA' lattices with plutonium dioxide fuel

    International Nuclear Information System (INIS)

    Halsall, M.J.

    1977-03-01

    A series of Westinghouse critical PuO 2 /UO 2 pin-cell assemblies is analysed using the lattice code WIMSD4. The results are presented in terms of computed k-effective values, with comment on the choice of method for calculating high leakage systems and on the adequacy of WIMSD4 for evaluating plutonium enriched lattices. (author)

  4. Westinghouse Hanford Company (WHC) standards/requirements identification document (S/RID)

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, G.L.

    1996-03-15

    This Standards/Requirements Identification Document (S/RID) set forth the Environmental Safety and Health (ES&H) standards/requirements for Westinghouse Hanford Company Level Programs, where implementation and compliance is the responsibility of these organizations. These standards/requirements are adequate to ensure the protection of the health and safety of workers, the public, and the environment.

  5. The AP600 advanced simplified nuclear power plant. Results of the test program and progress made toward final design approval

    Energy Technology Data Exchange (ETDEWEB)

    Bruschi, H.J. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

    1996-10-01

    At the 1994 Pacific Basin Conference, Mr. Bruschi presented a paper describing the AP600, Westinghouse`s advanced light water reactor design with passive safety features. Since then, a rigorous test program was completed and AP600 became the most thoroughly tested advanced reactor system design in history. Westinghouse is now well on its way toward receiving Final Design Approval from the U.S. Nuclear Regulatory Commission for AP600. In this paper, the results of the test program will be discussed and an update on prospects for building the plant will be covered. (author)

  6. Corporate science education: Westinghouse and the value of science in mid-twentieth century America.

    Science.gov (United States)

    Terzian, Sevan G; Shapiro, Leigh

    2015-02-01

    This study examines a largely neglected aspect of the history of science popularization in the United States: corporate depictions of the value of science to society. It delineates the Westinghouse Electric Corporation's portrayals of science to its shareholders, employees and consumers, and schoolchildren and educators during World War Two and the postwar era. Annual reports to shareholders, in-house news publications, publicity records, advertising campaigns, and educational pamphlets distributed to schools reveal the company's distinct, but complementary, messages for different stakeholders about the importance of science to American society. Collectively, Westinghouse encouraged these audiences to rely on scientists' expert leadership for their nation's security and material comforts. In an era of military mobilization, the company was able to claim that industry-led scientific research would fortify the nation and create unbounded prosperity. © The Author(s) 2013.

  7. Westinghouse-GOTHIC comparisons to AP600 passive containment cooling tests

    International Nuclear Information System (INIS)

    Kennedy, M.D.; Woodcock, J.; Gresham, J.A.

    1994-01-01

    Westinghouse-GOTHIC is a thermal-hydraulics code well suited to analyzing passively cooled containments which depend on heat removal primarily through the containment shell. The code includes boundary layer heat and mass transfer correlations. A liquid film convective energy transport model has been added to the Westinghouse-GOTHIC code to account for the sensible heat change of the applied exterior water. The objective of this paper is to compare the code's predictions of the AP600 large scale test facility with and without the liquid film convective energy transport model. The predicted vessel pressure and integrated heat rate with and without the film convective energy transport model will be compared to the measured data. (author)

  8. Westinghouse Hanford Company effluent releases and solid waste management report for 1987: 200/600/1100 Areas

    International Nuclear Information System (INIS)

    Coony, F.M.; Howe, D.B.; Voigt, L.J.

    1988-05-01

    The purpose of this report is to fulfill the reporting requirements of US Department of Energy (DOE) Order 5484.1, Environmental Protection, Safety, and Health Protection Information Reporting Requirements. Quantities of airborne and liquid wastes discharged by Westinghouse Hanford Company (Westinghouse Hanford) in the 200 Areas, 600 Area, and 1100 Area in 1987 are presented in this report. Also, quantities of solid wastes stored and buried by Westinghouse Hanford in the 200 Areas are presented in this report. The report is also intended to demonstrate compliance with Westinghouse Hanford administrative control limit (ACL) values for radioactive constituents and with applicable guidelines and standards for nonradioactive constituents. The summary of airborne release data, liquid discharge data, and solid waste management data for calendar year (CY) 1987 and CY 1986 are presented in Table ES-1. Data values for 1986 are cited in Table ES-1 to show differences in releases and waste quantities between 1986 and 1987. 19 refs., 3 figs., 19 tabs

  9. INEL design studies in support of the Westinghouse EPRI small plant study

    International Nuclear Information System (INIS)

    Burtt, J.D.; Kullberg, C.M.

    1986-03-01

    In support of the design effort of a Westinghouse EPRI small plant study, several analyses were performed at the Idaho National Engineering Laboratory. An analysis was performed to study fuel behavior under conditions of a limiting flow coastdown transient. Depressurization capabilities for the reactor coolant system were studied. The post-accident heat removal for the current containment design was studied. The results of all three studies are reported. 31 figs

  10. The role of Quality Oversight in nuclear and hazardous waste management and environmental restoration at Westinghouse Hanford Company

    International Nuclear Information System (INIS)

    Fouad, H.Y.

    1994-05-01

    The historical factors that led to the waste at Hanford are outlined. Westinghouse Hanford Company mission and organization are described. The role of the Quality Oversight organization in nuclear hazardous waste management and environmental restoration at Westinghouse Hanford Company is delineated. Tank Waste Remediation Systems activities and the role of the Quality Oversight organization are described as they apply to typical projects. Quality Oversight's role as the foundation for implementation of systems engineering and operation research principles is pointed out

  11. Generic risk insights for Westinghouse and Combustion Engineering pressurized water reactors

    International Nuclear Information System (INIS)

    Travis, R.; Taylor, J.; Fresco, A.; Chung, J.

    1990-11-01

    A methodology has been developed to extract generic risk-based information from probabilistic risk assessments (PRAs) of Westinghouse and Combustion Engineering (CE) pressurized water reactors (PWRs) and apply the insights gained to Westinghouse and Ce plants have not been subjected to a PRA. The available PRAs (five Westinghouse plants and one CE plant) were examined to identify the most probable, i.e., dominant accident sequences at each plant. The goal was to include all sequences which represented at least 80% of core damage frequency. If the same plant specific dominant accident sequence appeared within this boundary in at least two plant PRAs, the sequence was considered to be a representative sequence. Eleven sequences met this definition. From these sequences, the most important component failures and human errors that contributed to each sequence have been prioritized. Guidance is provided to prioritize the representative sequences and modify selected basic events that have been shown to be sensitive to the plant specific design or operating variations of the contributing PRAs. This risk-based guidance can be used for utility and NRC activities including operator training maintenance, design review, and inspections

  12. Aging mechanisms in the Westinghouse PWR [Pressurized Water Reactor] Control Rod Drive system

    International Nuclear Information System (INIS)

    Gunther, W.; Sullivan, K.

    1991-01-01

    An aging assessment of the Westinghouse Pressurized Water Reactor (PWR) Control Rod System (CRD) has been completed as part of the US NRC's Nuclear Plant Aging Research, (NPAR) Program. This study examined the design, construction, maintenance, and operation of the system to determine its potential for degradation as the plant ages. Selected results from this study are presented in this paper. The operating experience data were evaluated to identify the predominant failure modes, causes, and effects. From our evaluation of the data, coupled with an assessment of the materials of construction and the operating environment, we conclude that the Westinghouse CRD system is subject to degradation which, if unchecked, could affect its safety function as a plant ages. Ways to detect and mitigate the effects of aging are included in this paper. The current maintenance for the control rod drive system at fifteen Westinghouse PWRs was obtained through a survey conducted in cooperation with EPRI and NUMARC. The results of the survey indicate that some plants have modified the system, replaced components, or expanded preventive maintenance. Several of these activities have effectively addressed the aging issue. 2 refs., 2 figs., 2 tabs

  13. Advancing PWR fuel to meet customer needs

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, F W

    1987-03-01

    Since the introduction of the Optimized Fuel Assembly (OFA) for PWRs in the late 1970s, Westinghouse has continued to work with the utility customers to identify the greatest needs for further advance in fuel performance and reliability. The major customer requirements include longer fuel cycle at lower costs, increased fuel discharge burn-up, enhanced operating flexibility, all accompanied by even greater reliability. In response to these needs, Westinghouse developed Vantage 5 PWR fuel. To optimize reactor operations, Vantage 5 fuel features distinct advantages: integral fuel burnable absorbers, axial and radial blankets, intermediate flow mixers, a removable top nozzle, and assembly modifications to accommodate increased discharge burn-up.

  14. Survey report on the status of new energy in the U.S. On-site research centering on fuel cell, hydrogen energy, and wind energy (Westinghouse Electric Corporation); Beikoku shin energy jijo chosa hokokusho. Nenryo denchi, suiso furyoku energy wo chushin to suru jicchi chosa (Westinghouse Electric Corporation hen)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1982-02-01

    Under the auspices of the New Energy Foundation and the New Energy Industrial Forum technical development committee, a survey team is sent to the U.S. and conducts investigations there about fuel cells, hydrogen production, wind power generation, etc. Visited in the U.S. are the Advanced Energy System Division of the Westinghouse Electric Corporation. As for the phosphoric acid fuel cell, research and development is under way so that two 7.5MW demonstration plants will start service operation by 1987. As for the solid oxide fuel cell, a performance test has completed for a 15-cell model, and a life test is now under way. There is a plan to construct a 500kW plant in 1988. In the production of hydrogen by means of the sulfur hybrid decomposition process, a laboratory model with a capacity of 2L/min was built in 1978, and a life test is now under way for the constituent materials and catalysts. In the field of wind power, the Westinghouse Electric Corporation has developed a 200kW generator, which is now in operation in Mexico, Puerto Rico, Rhode Island, and Hawaii. (NEDO)

  15. Safety Evaluation Report related to the renewal of the operating license for the Westinghouse research reactor at Zion, Illinois (Docket No. 50-87)

    International Nuclear Information System (INIS)

    1984-09-01

    This Safety Evaluation Report, for the application filed by the Westinghouse Electric Company, for renewal of operating license number R-119 to continue to operate the research reactor, has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is operated by Westinghouse and is located in Zion, Illinois. The staff concludes that the reactor facility can continue to be operated by Westinghouse without endangering the health and safety of the public

  16. Recent advances in fuel product and manufacturing process development

    International Nuclear Information System (INIS)

    Slember, R.J.; Doshi, P.K.

    1987-01-01

    This paper discusses advancements in commercial nuclear fuel products and manufacturing made by the Westinghouse Electric Corporation in response to the commercial nuclear fuel industry's demand for high reliability, increased plant availability and improved operating flexibility. The features and benefits of Westinghouse's most advanced fuel products--VANTAGE 5 for PWR plants and QUAD+ for BWR plants--are described, as well as 'high performance' fuel concepts now under development for delivery in the late 1980s. The paper also disusses the importance of in-process quality control throughout manufacturing towards reducing product variability and improving fuel reliability. (author)

  17. 10 CFR Appendix D to Part 52 - Design Certification Rule for the AP1000 Design

    Science.gov (United States)

    2010-01-01

    ... severe accident mitigation design alternatives in appendix 1B of the generic DCD are not part of this... issues concerning severe accident mitigation design alternatives associated with the information in the... appendix whose site parameters are within those specified in the severe accident mitigation design...

  18. The research on AP1000 nuclear main pumps’ complete characteristics and the normalization method

    International Nuclear Information System (INIS)

    Zhu, Rongsheng; Liu, Yong; Wang, Xiuli; Fu, Qiang; Yang, Ailing; Long, Yun

    2017-01-01

    Highlights: • Complete characteristics of main pump are researched into. • The quadratic character of head and torque under some operatings. • The characteristics tend to be the same under certain conditions. • The normalization method gives proper estimations on external characteristics. • The normalization method can efficiently improve the security computing. - Abstract: The paper summarizes the complete characteristics of nuclear main pumps based on experimental results and makes a detailed study, and then draws a series of important conclusions: with regard to the overall flow area, the runaway operating and 0-revolving-speed operating of nuclear main pumps both have quadratic characteristics; with regard to the infinite flow, the braking operation and the 0-revolving-speed operation show consistent external characteristics. To remedy the shortcomings of the traditional complete-characteristic expression with regards to only describing limited flow sections at specific revolving speeds, the paper proposes a normalization method. As an important boundary condition of the security computing of unstable transient process of the primary reactor coolant pump and the nuclear island primary circuit and secondary circuit, the precision of complete-characteristic data and curve impacts the precision of security computing. A normalization curve obtained by applying the normalization method to process complete-characteristic data could correctly, completely and precisely express the complete characteristics of the primary reactor coolant pump under any rotational speed and full flow, and is capable of giving proper estimations on external characteristics of the flow outside the test range and even of the infinite flow. These advantages are of great significance for the improvement of security computing of transient processes of the primary reactor coolant pump and the circuit system.

  19. Pneumatic transport system development: residuals and releases program at Westinghouse Cheswick site

    International Nuclear Information System (INIS)

    Larouere, P.J.; Shoulders, J.L.

    1979-01-01

    Plutonium oxide and uranium oxide powders are processed within glove boxes or within confinement systems during the fabrication of mixed oxide (MOX) pellets for recycle fuel. The release of these powders to the glove box or to the confinement results in some airborne material that is deposited in the enclosure or is carried in the air streams to the effluent air filtration system. Release tests on simulated leaks in pneumatic transport equipment and release tests on simulated failures with powder blending equipment were conducted. A task to develop pneumatic transport for the movement of powders within an MOX fabrication plant has been underway at the Westinghouse Research Laboratories. While testing and evaluating selected pneumatic transport components on a full scale were in progress, it was deemed necessary that final verification of the technology would have to be performed with plutonium-bearing powders because of the marked differences in certain properties of plutonium from those of uranium oxides. A smaller was designed and constructed for the planned installation in glove boxes at the Westinghouse Plutonium Fuel Development Laboratory. However, prior to use with plutonium it was agreed that this system be set up and tested with uranium oxide powder. The test program conducted at the Westinghouse Cheswick site was divided into two major parts. The first of these examined the residuals left as a result of the pneumatic transport of nuclear fuel powders and verified the operability of this one-third scale system. The second part of the program studied the amount of powder released to the air when off-standard process procedures or maintenance operations were conducted on the pneumatic transport system. Air samplers located within the walk-in box housing the transport loop were used to measure the solids concentration in the air. From this information, the total amount of airborne powder was determined

  20. Westinghouse Hanford Company environmental surveillance annual report -- 200/600 Areas

    International Nuclear Information System (INIS)

    Schmidt, J.W.; Huckfeldt, C.R.; Johnson, A.R.; McKinney, S.M.

    1990-06-01

    This document presents the results of near-field environmental surveillance as performed by Westinghouse Hanford Company in 1989 for the Operations Area of the Hanford Site, Richland, Washington. These activities were conducted in the 200 and 600 Areas to assess operational control on the work environment. Surveillance activities included external radiation measurements and radiological surveys of waste disposal sites, radiological control areas, and roads, as well as sampling and analysis of ambient air, surface water, groundwater, sediments, soil, and biota. 15 refs., 3 figs., 1 tab

  1. Effects of natural phenomena on the Westinghouse Electric Corporation Plutonium Fuels Development Laboratory at Cheswick, Pennsylvania

    International Nuclear Information System (INIS)

    1979-11-01

    One aim of the analysis is to examine the plant with the objective of improving its ability to withstand adverse natural phenomena without loss of capability to protect the public. The relatively small risk to the public from the unlikely events discussed (earthquake, flood, tornado) would indicate that the public is not seriously threatened by the presence of the Westinghouse PFDL. Thus, it is the judgment of the staff that the benefits to be gained by substantial plant improvements to further mitigate against adverse natural phenomena are not cost effective

  2. Effects of the reactor coolant pumps following a small break in a Westinghouse PWR

    International Nuclear Information System (INIS)

    Koenig, J.E.

    1983-10-01

    Numerical simulations of the thermal-hydraulic events following a small cold-leg break in a Westinghouse pressurized water reactor were performed to address the pumps-on/off issue. The mode of pump operation was varied in each calculation to ascertain the optimum mode. It was found that pump operation was not critical for this break size and location because the fuel rods remained cool in all accidents analyzed. In terms of system mass, however, it was preferable to leave the pumps in operation

  3. Detection and mitigating rod drive control system degradation in Westinghouse PWRs

    International Nuclear Information System (INIS)

    Gunther, W.; Sullivan, K.

    1990-01-01

    A study of the effects of aging on the Westinghouse Control Rod Drive (CRD) System was performed as part of the US NRC's Nuclear Plant aging Research (NPAR) Program. For the study, the CRD system boundary includes the power and logic cabinets associated with the manual control rod movement, and the control rod mechanism itself. The aging-related degradation of the interconnecting cables and connectors and the rod position indicating system also were considered. This paper presents the results of that study pertaining to the electrical and instrumentation portions of the CRD system including ways to detect and mitigate system degradation

  4. Westinghouse Hanford Company Environmental surveillance annual report--200/600 Areas

    International Nuclear Information System (INIS)

    Schmidt, J.W.; Huckfeldt, C.R.; Johnson, A.R.; McKinney, S.M.

    1991-06-01

    This document presents the results of near-field environmental surveillance in 1990 of the Operations Area of the Hanford Site, in south central Washington State, as performed by Westinghouse Hanford Company. These activities are conducted in the 200 and 600 Areas to assess and control the impacts of operations on the workers and the local environment. Surveillance activities include sampling and analyses of ambient air, surface water, groundwater, sediments, soil, and biota. Also, external radiation measurements and radiological surveys are taken of waste disposal sites, radiological control areas, and roads. 16 refs., 3 figs., 1 tab

  5. Evaluation of selected parameters on exposure rates in Westinghouse designed nuclear power plants

    International Nuclear Information System (INIS)

    Bergmann, C.A.

    1989-01-01

    During the past ten years, Westinghouse under EPRI contract and independently, has performed research and evaluation of plant data to define the trends of ex-core component exposure rates and the effects of various parameters on the exposure rates. The effects of the parameters were evaluated using comparative analyses or empirical techniques. This paper updates the information presented at the Fourth Bournemouth Conference and the conclusions obtained from the effects of selected parameters namely, coolant chemistry, physical changes, use of enriched boric acid, and cobalt input on plant exposure rates. The trends of exposure rates and relationship to doses is also presented. (author)

  6. WESTINGHOUSE 17X17 MOX PWR ASSEMBLY - WASTE PACKAGE CRITICALITY ANALYSIS (SCPB: N/A)

    International Nuclear Information System (INIS)

    J.W. Davis

    1996-01-01

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to compare the criticality potential of Westinghouse 17 x 17 mixed oxide (MOX) PWR fuel with the Design Basis spent nuclear fuel (SNF) analyzed previously (Ref. 5.1, 5.2). The basis of comparison will be the conceptual design Multi-Purpose Canister (MPC) PWR waste package concepts. The objectives of this evaluation are to show that the criticality potential of the MOX fuel is equal to or lower than the DBF or, if necessary, indicate what additional measures are required to make it so

  7. A consortium approach to commercialized Westinghouse solid oxide fuel cell technology

    Science.gov (United States)

    Casanova, Allan

    Westinghouse is developing its tubular solid oxide fuel cells (SOFCs) for a variety of applications in stationary power generation markets. By pressurizing a SOFC and integrating it with a gas turbine (GT), power systems with efficiencies as high as 70-75% can be obtained. The first such system will be tested in 1998. Because of their extraordinarily high efficiency (60-70%) even in small sizes the first SOFC products to be offered are expected to be integrated SOFC/GT power systems in the 1-7 MW range, for use in the emerging distributed generation (DG) market segment. Expansion into larger sizes will follow later. Because of their modularity, environmental friendliness and expected cost effectiveness, and because of a worldwide thrust towards utility deregulation, a ready market is forecasted for baseload distributed generation. Assuming Westinghouse can complete its technology development and reach its cost targets, the integrated SOFC/GT power system is seen as a product with tremendous potential in the emerging distributed generation market. While Westinghouse has been a leader in the development of power generation technology for over a century, it does not plan to manufacture small gas turbines. However, GTs small enough to integrate with SOFCs and address the 1-7 MW market are generally available from various manufacturers. Westinghouse will need access to a new set of customers as it brings baseload plants to the present small market mix of emergency and peaking power applications. Small cogeneration applications, already strong in some parts of the world, are also gaining ground everywhere. Small GT manufacturers already serve this market, and alliances and partnerships can enhance SOFC commercialization. Utilities also serve the DG market, especially those that have set up energy service companies and seek to grow beyond the legal and geographical confines of their current regulated business. Because fuel cells in general are a new product, because small

  8. Sensitivity Analysis of Onsite Atmospheric Dispersion Factor in Westinghouse type NPP in KOREA

    International Nuclear Information System (INIS)

    Lee, Seung Chan; Yoon, Duk Joo; Song, Dong Soo

    2016-01-01

    ARCON96 is a NRC licensed air dispersion model to evaluate onsite atmospheric relative concentration X/Q. The purpose of this paper is to provide some results for checking and testing the functionalities of ARCON96. Specially, this code is optimized to estimate a habitability of control room. Since NUREG 0737 issue, the control room habitability has been studied for a FSAR (Final Safety Analysis Report). Some assumptions and methodology is used in this paper. Some methodology is introduced in this paper. The reason of the selection of 2-loop Westinghouse NPP is because of carrying out the study project for the 2-loop Westinghouse NPP in the condition of the defueled NPP condition. Onsite atmospheric dispersion factor sensitivity is performed. Key impact factor is reviewed. Some results are below: a. Time averaged effect of X/Q is timely increased. b. ARCON96 code is more conservative at the low wind speed conditions. c. Building wake impact is significant in the condition of unstable atmospheric class with more than 7m/sec of wind speed. d. Plume meander effect is strong when the distance from the release point is small. e. The other plume meander effect is strong when the meander duration time is accumulated Finally, these results show that the appropriate conservation of ARCON96 is appeared in some conditions. Also these results seem to be in good agreement with NRC Regulatory Guide and positions

  9. Sensitivity Analysis of Onsite Atmospheric Dispersion Factor in Westinghouse type NPP in KOREA

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seung Chan; Yoon, Duk Joo; Song, Dong Soo [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    ARCON96 is a NRC licensed air dispersion model to evaluate onsite atmospheric relative concentration X/Q. The purpose of this paper is to provide some results for checking and testing the functionalities of ARCON96. Specially, this code is optimized to estimate a habitability of control room. Since NUREG 0737 issue, the control room habitability has been studied for a FSAR (Final Safety Analysis Report). Some assumptions and methodology is used in this paper. Some methodology is introduced in this paper. The reason of the selection of 2-loop Westinghouse NPP is because of carrying out the study project for the 2-loop Westinghouse NPP in the condition of the defueled NPP condition. Onsite atmospheric dispersion factor sensitivity is performed. Key impact factor is reviewed. Some results are below: a. Time averaged effect of X/Q is timely increased. b. ARCON96 code is more conservative at the low wind speed conditions. c. Building wake impact is significant in the condition of unstable atmospheric class with more than 7m/sec of wind speed. d. Plume meander effect is strong when the distance from the release point is small. e. The other plume meander effect is strong when the meander duration time is accumulated Finally, these results show that the appropriate conservation of ARCON96 is appeared in some conditions. Also these results seem to be in good agreement with NRC Regulatory Guide and positions.

  10. Standard technical specifications, Westinghouse Plants: Bases (Sections 3.4--3.9). Volume 3, Revision 1

    International Nuclear Information System (INIS)

    1995-04-01

    This NUREG contains the improved Standard Technical Specifications (STS) for Westinghouse plants. Revision 1 incorporates the cumulative changes to Revision 0, which was published in September 1992. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, specifically the Westinghouse Owners Group (WOG), NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993 (58 FR 39132). Licensees are encouraged to upgrade their technical specifications consistent with those criteria and conforming, to the extent practical and consistent with the licensing basis for the facility, to Revision 1 to the improved STS. The Commission continues to place the highest priority on requests for complete conversions to the improved STS. Licensees adopting portions of the improved STS to existing technical specifications should adopt all related requirements, as applicable, to achieve a high degree of standardization and consistency

  11. Advantages of Westinghouse BWR control rod drop accidents methodology utilizing integrated POLCA-T code

    International Nuclear Information System (INIS)

    Panayotov, Dobromir

    2008-01-01

    The paper focuses on the activities pursued by Westinghouse in the development and licensing of POLCA-T code Control Rod Drop Accident (CRDA) Methodology. The comprehensive CRDA methodology that utilizes PHOENIX4/POLCA7/POLCA-T calculation chain foresees complete cycle-specific analysis. The methodology consists of determination of candidates of control rods (CR) that could cause a significant reactivity excursion if dropped throughout the entire fuel cycle, selection of limiting initial conditions for CRDA transient simulation and transient simulation itself. The Westinghouse methodology utilizes state-of-the-art methods. Unnecessary conservatisms in the methodology have been avoided to allow the accurate prediction of margin to design bases. This is mainly achieved by using the POLCA-T code for dynamic CRDA evaluations. The code belongs to the same calculation chain that is used for core design. Thus the very same reactor, core, cycle and fuel data base is used. This allows also reducing the uncertainties of input data and parameters that determine the energy deposition in the fuel. Uncertainty treatment, very selective use of conservatisms, selection of the initial conditions for limiting case analyses, incorporation into POLCA-T code models of the licensed fuel performance code are also among the means of performing realistic CRDA transient analyses. (author)

  12. Standard technical specifications, Westinghouse Plants: Bases (Sections 2.0--3.3). Volume 2, Revision 1

    International Nuclear Information System (INIS)

    1995-04-01

    This NUREG contains the improved Standard Technical Specifications (STS) for Westinghouse plants. Revision 1 incorporates the cumulative changes to Revision 0, which was published in September 1992. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, specifically the Westinghouse Owners Group (WOG), NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993 (58 FR 39132). Licensees are encouraged to upgrade their technical specifications consistent with those criteria and conforming, to the extent practical and consistent with the licensing basis for the facility, to Revision 1 to the improved STS. The Commission continues to place the highest priority on requests for complete conversions to the improved STS. Licensees adopting portions of the improved STS to existing technical specifications should adopt all related requirements, as applicable, to achieve a high degree of standardization and consistency

  13. Effects of RCP trip when recovering HPSI during LOCA in a Westinghouse PWR

    Energy Technology Data Exchange (ETDEWEB)

    Montero-Mayorga, Javier, E-mail: fj.montero@alumnos.upm.es; Queral, César; Rivas-Lewicky, Julio; González-Cadelo, Juan

    2014-12-15

    Highlights: • If HPSI is recovered during SBLOCA and RCPs are tripped core damage can be reached. • If the RCPs are tripped once the accumulators have injected the damage can be avoided. • If only 2 out of 3 RCPs are tripped the damage can be also avoided. • Improvements are proposed to the EOPs in order to avoid possible damage. - Abstract: Current Westinghouse Emergency Operating Procedures (EOPs) indicate initially that the operator must keep the reactor coolant pumps (RCPs) running during a Small Break Loss of Coolant Accident (SBLOCA) if there is unavailability of high pressure safety injection (HPSI) system in order to cool the core by forced convection. However, the crew must follow different EOPs along the transient depending on its evolution. In these EOPs there are several conditions which indicate the necessity of tripping one or more RCPs when HPSI is recovered. In this paper the occurrence of a SBLOCA with unavailability of HPSI has been analyzed with a model of Almaraz Nuclear Power Plant (Westinghouse 3 Loop) for TRACE code V5.0 patch 1. Two different approaches have been considered: the first one, taking into account Optimal Recovery Guidelines (ORGs) and in the second approach, the transition to Function Restoration Guidelines (FRGs) due to inadequate core cooling (ICC) conditions is considered. Results of this paper lead to the implementation of an improvement in current EOPs regarding how many RCPs should be tripped during SBLOCA sequences.

  14. Westinghouse power distribution monitoring experience at Duke Power's McGuire Unit 1

    International Nuclear Information System (INIS)

    Grobmyer, L.R.; Cash, M.T.; Kitlan, M.S.; Impink, A.J. Jr.

    1987-01-01

    In the evolution of the Westinghouse methodology of assuring safe core power distributions, emphasis was placed on analysis and not on continuous detailed core monitoring. Power distribution monitoring is currently achieved by periodic surveillances using the movable in-core detector system (MIDS) and by continuous observations of the two-section excore power range detectors. Control of the power distribution is regulated by limits on the indications from these systems, by limits on control rod insertion, and by operational constraints on the position indication systems. As more plants come on line and as more utilities take over the fuel design function for themselves, the desire for better core monitoring becomes evident. Also, the need and desire by the utilities to have more control over their operating margin has motivated the industry to offer and/or upgrade core monitoring systems. Westinghouse and Duke Power are participants in a joint development program to finalize the development of the core on-line surveillance monitoring and operations system (COSMOS). This final stage of development consists of prototype field trials at the McGuire Nuclear Plant. The purpose of the prototype program is to determine how well the design objectives are met and how to improve the system based on the operating experience at McGuire. Another purpose of this prototype program is to generate the necessary experience and information to develop a topical report for the US Nuclear Regulatory Commission to obtain a licensing basis for technical specification relaxation

  15. Westinghouse experience over the past 10 years in negotiating and constructing nuclear power plants

    International Nuclear Information System (INIS)

    Richards, D.E.

    1979-01-01

    Reason for delays in delivery times for nuclear plant are discussed in the light of Westinghouse experience. Today the lead time for the construction of the plant is no longer dictated by the lead time of the nuclear steam supply system. The increased complexity of contract negotiations and of standards and specifications contributes to the delays. Site work is constantly subject to delays due to various labour problems. The main delays stem from regulatory authorities, environmentalists and political considerations. Lateness on the plant causes problems of warranty, storage of equipment and of finance. Westinghouse procedures for alleviating delays during erection are outlined. As the start-up schedule dictates erection, purchasing and design, it should be established as early as possible. A typical overall schedule for a PWR is outlined. It is concluded that completion of plant within schedule requires decisions on basic principles and sufficient detailed planning and organisational structures to be established before the start of the project followed by strong project management. The discussion following the conference is also recorded. (U.K.)

  16. Experimental prediction of tube support interaction characteristics in steam generators: Volume 2, Westinghouse Model 51 flow entrance region: Topical report

    International Nuclear Information System (INIS)

    Haslinger, K.H.

    1988-06-01

    Tube-to-tube support interaction characterisitics were determined experimentally on a single tube, multi-span geometry, representative of the Westinghouse Model 51 steam generator economizer design. Results, in part, became input for an autoclave type wear test program on steam generator tubes, performed by Kraftwerk Union (KWU). More importantly, the test data reported here have been used to validate two analytical wear prediction codes; the WECAN code, which was developed by Westinghouse, and the ABAQUS code which has been enhanced for EPRI by Foster Wheeler to enable simulation of gap conditions (including fluid film effects) for various support geometries

  17. Superfund record of decision (EPA Region 3), Westinghouse Elevator Company Plant, Operable Unit 2, Cumberland Township, Adams County, Gettysburg, PA, March 31, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-04-01

    This Record of Decision (ROD) presents the selected remedial action for Operable Unit 2 (Soils) at the Westinghouse Elevator Company Plant Site in Adams County, Pennsylvania. The selected remedy for the soils at the Westinghouse Elevator Plant is No Additional Action for this Operable Unit. The other alternatives evaluated would produce little or no environmental benefit at substantial cost.

  18. Solution of closing of the columns of thermocouples in Asco reactors 1 and 2 with Cetna of Westinghouse; Solucion de cierre de las columnas de termopares en reactores Asco 1 and 2 con Cetna de Westinghouse

    Energy Technology Data Exchange (ETDEWEB)

    Sunjic, B.; Reichenbach, M.; Llibre, E.

    2014-10-01

    Occasionally, small leaks have been discovered in operating PWRs in the Thermo Couple columns Penetrations. In order to mitigate this issue, Westinghouse has designed and developed the CETNA element, which does not use cono-seals. This article shows the CETNA supply for Asco NPP to prevent potential leaks in the penetrations. (Author)

  19. "A Highly Selected Strain of Guinea Pigs": The Westinghouse Science Talent Search and Educational Meritocracy, 1942-1958

    Science.gov (United States)

    Terzian, Sevan G.; Rury, John L.

    2014-01-01

    Overview: This article examines the Westinghouse Science Talent Search over the first sixteen years of its operation. A national contest involving thousands of high school seniors annually, it reflected a growing national concern with developing scientific manpower in the midst of global conflict, the Cold War, and a growing military-industrial…

  20. Demonstration of retrieval methods for Westinghouse Hanford Corporation October 20, 1995

    International Nuclear Information System (INIS)

    1996-10-01

    Westinghouse Hanford Corporation has been pursuing strategies to break up and retrieve the radioactive waste material in single shell storage tanks at the Hanford Nuclear Reservation, by working with non-radioactive ''saltcake'' and sludge material that simulate the actual waste. It has been suggested that the use of higher volumes of water than used in the past (10 gpm nozzles at 10,000 psi) might be successful in breaking down the hard waste simulants. Additionally, the application of these higher volumes of water might successfully be applied through commercially available tooling using methods similar to those used in the deslagging of large utility boilers. NMW Industrial Services, Inc., has proposed a trial consisting of three approaches each to dislodging both the solid (saltcake) simulant and the sludge simulant

  1. Demonstration of retrieval methods for Westinghouse Hanford Corporation October 20, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-10-01

    Westinghouse Hanford Corporation has been pursuing strategies to break up and retrieve the radioactive waste material in single shell storage tanks at the Hanford Nuclear Reservation, by working with non-radioactive ``saltcake`` and sludge material that simulate the actual waste. It has been suggested that the use of higher volumes of water than used in the past (10 gpm nozzles at 10,000 psi) might be successful in breaking down the hard waste simulants. Additionally, the application of these higher volumes of water might successfully be applied through commercially available tooling using methods similar to those used in the deslagging of large utility boilers. NMW Industrial Services, Inc., has proposed a trial consisting of three approaches each to dislodging both the solid (saltcake) simulant and the sludge simulant.

  2. Mixcore safety analysis approach used for introduction of Westinghouse fuel assemblies in Ukraine

    International Nuclear Information System (INIS)

    Abdullayev, A.; Baidullin, V.; Maryochin, A.; Sleptsov, S.; Kulish, G.

    2008-01-01

    Six Westinghouse Lead Test Assemblies (LTA) were installed in 2005 and are currently operated in Unit 3 of the South Ukraine NPP (SUNPP) under the Ukraine Nuclear Fuel Qualification Project. At the early stages of the LTAs implementation in Ukraine, there was no experience of licensing of new fuel types, which explains the need to develop approaches for safety substantiation of LTAs. This presentation considers some approaches for performing of safety analysis of the design basis Initiating Events (IE) for the LTA fuel cycles. These approaches are non-standard in terms of the established practices for obtaining the regulatory authorities' permission for the core operation. The analysis was based on the results of the FA and reactor core thermal hydraulic and nuclear design

  3. Assessment of ISLOCA risk: Methodology and application to a Westinghouse four-loop ice condenser plant

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, D.L.; Auflick, J.L.; Haney, L.N. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1992-04-01

    Inter-system loss-of-coolant accidents (ISLOCAs) have been identified as important contributors to offsite risk for some nuclear power plants. A methodology has been developed for identifying and evaluating plant-specific hardware designs, human factors issues, and accident consequence factors relevant to the estimation of ISLOCA core damage frequency and risk. This report presents a detailed description of the application of this analysis methodology to a Westinghouse four-loop ice condenser plant. This document also includes appendices A through I which provide: System descriptions; ISLOCA event trees; human reliability analysis; thermal hydraulic analysis; core uncovery timing calculations; calculation of system rupture probability; ISLOCA consequences analysis; uncertainty analysis; and component failure analysis.

  4. cmpXLatt: Westinghouse automated testing tool for nodal cross section models

    International Nuclear Information System (INIS)

    Guimaraes, Petri Forslund; Rönnberg, Kristian

    2011-01-01

    The procedure for evaluating the merits of different nodal cross section representation models is normally both cumbersome and time consuming, and includes many manual steps when preparing appropriate benchmark problems. Therefore, a computer tool called cmpXLatt has been developed at Westinghouse in order to facilitate the process of performing comparisons between nodal diffusion theory results and corresponding transport theory results on a single node basis. Due to the large number of state points that can be evaluated by cmpXLatt, a systematic and comprehensive way of performing verification and validation of nodal cross section models is provided. This paper presents the main features of cmpXLatt and demonstrates the benefits of using cmpXLatt in a real life application. (author)

  5. Assessment of ISLOCA risk: Methodology and application to a Westinghouse four-loop ice condenser plant

    International Nuclear Information System (INIS)

    Kelly, D.L.; Auflick, J.L.; Haney, L.N.

    1992-04-01

    Inter-system loss-of-coolant accidents (ISLOCAs) have been identified as important contributors to offsite risk for some nuclear power plants. A methodology has been developed for identifying and evaluating plant-specific hardware designs, human factors issues, and accident consequence factors relevant to the estimation of ISLOCA core damage frequency and risk. This report presents a detailed description of the application of this analysis methodology to a Westinghouse four-loop ice condenser plant. This document also includes appendices A through I which provide: System descriptions; ISLOCA event trees; human reliability analysis; thermal hydraulic analysis; core uncovery timing calculations; calculation of system rupture probability; ISLOCA consequences analysis; uncertainty analysis; and component failure analysis

  6. TPDWR2: thermal power determination for Westinghouse reactors, Version 2. User's guide

    International Nuclear Information System (INIS)

    Kaczynski, G.M.; Woodruff, R.W.

    1985-12-01

    TPDWR2 is a computer program which was developed to determine the amount of thermal power generated by any Westinghouse nuclear power plant. From system conditions, TPDWR2 calculates enthalpies of water and steam and the power transferred to or from various components in the reactor coolant system and to or from the chemical and volume control system. From these results and assuming that the reactor core is operating at constant power and is at thermal equilibrium, TPDWR2 calculates the thermal power generated by the reactor core. TPDWR2 runs on the IBM PC and XT computers when IBM Personal Computer DOS, Version 2.00 or 2.10, and IBM Personal Computer Basic, Version D2.00 or D2.10, are stored on the same diskette with TPDWR2

  7. Westinghouse Hanford Company operational environmental monitoring annual report, calendar year 1994

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, J.; Fassett, J.W.; Johnson, A.R.; Johnson, V.G.; Markes, B.M.; McKinney, S.M.; Moss, K.J.; Perkins, C.J.; Richterich, L.R.

    1995-08-01

    This document presents the results of the Westinghouse Hanford Company near-facility operational environmental monitoring for 1994 in the 100, 200/600, and 300/400 Areas of the Hanford Site, in south-central Washington State. Surveillance activities included sampling and analyses of ambient air surface water, groundwater, soil, sediments, and biota. Also, external radiation measurements and radiological surveys were taken at waste disposal sites, radiologically controlled areas, and roads. These activities were conducted to assess and control the effects of nuclear facilities and waste sites on the local environment. In addition, diffuse sources were monitored to determine compliance with Federal, State, and/or local regulations. In general, although effects from nuclear facilities are still seen on the Hanford Site and radiation levels are slightly elevated when compared to offsite locations, the differences are less than in previous years.

  8. Westinghouse Hanford Company operational environmental monitoring annual report - calendar year 1995

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, J.W., Westinghouse Hanford

    1996-07-30

    This document summarizes the results of the Westinghouse Hanford Company (WHC) near-facility operational environmental monitoring for 1995 in the 100, 200/600, and 300/400 Areas of the Hanford Site, in south-central Washington State. Surveillance activities included sampling and analyses of ambient air, surface water,groundwater, soil, sediments, and biota. Also, external radiation measurements and radiological surveys were taken at waste disposal sites, radiologically controlled areas, and roads. These activities were conducted to assess and control the effects of nuclear facilities and waste sites on the local environment. In addition, diffuse sources were monitored to determine compliance with Federal, State, and/or local regulations. In general, although effects from nuclear facilities can still be observed on the Hanford Site and radiation levels are slightly elevated when compared to offsite locations, the differences are less than in previous years.

  9. Westinghouse Electric Company experiences in chemistry on-line monitoring in Eastern European nuclear power plants

    International Nuclear Information System (INIS)

    Balavage, J.

    2001-01-01

    Westinghouse Electric Company has provided a number of Chemistry On-Line Monitoring (OLM) Systems to Nuclear Power Plants in Eastern Europe. Eleven systems were provided to the Temelin Nuclear Power Plant in the south of the Czech Republic. Four systems were provided to the Russian NPP at Novovoronezh. In addition, a system design was developed for primary side chemistry monitoring for units 5 and 6 of another eastern European VVER. The status of the Temelin OLM systems is discussed including updates to the Temelin designs, and the other Eastern European installations and designs are also described briefly. Some of the problems encountered and lessons learned from these projects are also discussed. (R.P.)

  10. Application of quality assurance to scientific activities at Westinghouse Hanford Company

    International Nuclear Information System (INIS)

    Delvin, W.L.; Farwick, D.G.

    1988-01-01

    The application of quality assurance to scientific activities has been an ongoing subject of review, discussion, interpretation, and evaluation within the nuclear community for the past several years. This paper provides a discussion on the natures of science and quality assurance and presents suggestions for integrating the two successfully. The paper shows how those actions were used at the Westinghouse Hanford Company to successfully apply quality assurance to experimental studies and materials testing and evaluation activities that supported a major project. An important factor in developing and implementing the quality assurance program was the close working relationship that existed between the assigned quality engineers and the scientists. The quality engineers, who had had working experience in the scientific disciplines involved, were able to bridge across from the scientists to the more traditional quality assurance personnel who had overall responsibility for the project's quality assurance program

  11. Environmental programs for grades K-12 sponsored by the Westinghouse Waste Isolation Division Educational Programs Department

    International Nuclear Information System (INIS)

    Mikel, C.J.

    1993-01-01

    The Waste Isolation Pilot Plant (WIPP) created its educational programs department in 1990 as a result of the Secretary of Energy's focus on education stated in SEN-23-90. This Secretary of Energy Notice reflects the focus for US Department of Energy facilities to enhance education through their resources (both human and financial) with an emphasis on math and science. The mission of the Westinghouse Waste Isolation Division (WID) educational programs department is to enhance education at all levels and to promote educational experiences that give students the opportunity to make decisions and develop skills for productive lives. Programs have been developed around the environmental monitoring department, to give students from different grade levels hands on experiences in the environmental sciences field to stimulate their interest in the natural sciences

  12. Selection and development of advanced nuclear fuel products

    International Nuclear Information System (INIS)

    Stucker, David L.; Miller, Richard S.; Arnsberger, Peter L.

    2004-01-01

    The highly competitive international marketplace requires a continuing product development commitment, short development cycle times and timely, on-target product development to assure customer satisfaction and continuing business. Westinghouse has maintained its leadership position within the nuclear fuel industry with continuous developments and improvements to fuel assembly materials and design. This paper presents a discussion of the processes used by Westinghouse in the selection and refinement of advanced concepts for deployment in the highly competitive US and international nuclear fuel fabrication marketplace. (author)

  13. VVANTAGE 6 - an advanced fuel assembly design for VVER reactors

    International Nuclear Information System (INIS)

    Doshi, P.K.; DeMario, E.E.; Knott, R.P.

    1993-01-01

    Over the last 25 years, Westinghouse fuel assemblies for pressurized water reactors (PWR's) have undergone significant changes to the current VANTAGE 5. VANTAGE 5 PWR fuel includes features such as removable top nozzles, debris filter bottom nozzles, low-pressure-drop zircaloy grids, zircaloy intermediate flow mixing grids, optimized fuel rods, in-fuel burnable absorbers, and increased burnup capability to region average values of 48000 MWD/MTU. These features have now been adopted to the VVER reactors. Westinghouse has completed conceptual designs for an advanced fuel assembly and other core components for VVER-1000 reactors known as VANTAGE 6. This report describes the VVANTAGE 6 fuel assembly design

  14. Westinghouse-GOTHIC distributed parameter modelling for HDR test E11.2

    International Nuclear Information System (INIS)

    Narula, J.S.; Woodcock, J.

    1994-01-01

    The Westinghouse-GOTHIC (WGOTHIC) code is a sophisticated mathematical computer code designed specifically for the thermal hydraulic analysis of nuclear power plant containment and auxiliary buildings. The code is capable of sophisticated flow analysis via the solution of mass, momentum, and energy conservation equations. Westinghouse has investigated the use of subdivided noding to model the flow patterns of hydrogen following its release into a containment atmosphere. For the investigation, several simple models were constructed to represent a scale similar to the German HDR containment. The calculational models were simplified to test the basic capability of the plume modeling methods to predict stratification while minimizing the number of parameters. A large empty volume was modeled, with the same volume and height as HDR. A scenario was selected that would be expected to stably stratify, and the effects of noding on the prediction of stratification was studied. A single phase hot gas was injected into the volume at a height similar to that of HDR test E11.2, and there were no heat sinks modeled. Helium was released into the calculational models, and the resulting flow patterns were judged relative to the expected results. For each model, only the number of subdivisions within the containment volume was varied. The results of the investigation of noding schemes has provided evidence of the capability of subdivided (distributed parameter) noding. The results also showed that highly inaccurate flow patterns could be obtained by using an insufficient number of subdivided nodes. This presents a significant challenge to the containment analyst, who must weigh the benefits of increased noding with the penalties the noding may incur on computational efficiency. Clearly, however, an incorrect noding choice may yield erroneous results even if great care has been taken in modeling accurately all other characteristics of containments. (author). 9 refs., 9 figs

  15. Safety evaluation report on Westinghouse Electric Company ECCS evaluation model for plants equipped with upper head injection

    International Nuclear Information System (INIS)

    Lauben, G.N.; Wagner, N.H.; Israel, S.L.; McPherson, G.D.; Hodges, M.W.

    1978-04-01

    For plants which include an ice condenser containment concept, Westinghouse has planned an additional safety system known as the upper head injection (UHI) system to augment the emergency core cooling system. This system is comprised of additional accumulator tanks and piping arranged to supply cooling water to the top of the core during the blowdown period following a postulated large-break loss-of-coolant accident (LOCA). The objective of UHI is to add to the core cooling provided by the conventional emergency core cooling system (ECCS) and so permit operation at linear heat rates comparable to those permitted in plants utilizing the dry containment concept. In this way, plants which include the UHI system would have greater operating flexibility while still meeting the acceptance criteria as defined in paragraph 50.46 of 10 CFR Part 50. This review is concerned with those changes to the Westinghouse ECCS evaluation model that have been proposed for the UHI-LOCA model

  16. Core damage frequency prespectives for BWR 3/4 and Westinghouse 4-loop plants based on IPE results

    International Nuclear Information System (INIS)

    Dingman, S.; Camp, S.; LaChance, J.; Mary Drouin

    1995-01-01

    This paper discusses the core damage frequency (CDF) insights gained by analyzing the results of the Individual Plant Examinations (IPES) for two groups of plants: boiling water reactor (BWR) 3/4 plants with Reactor Core Isolation Cooling systems, and Westinghouse 4-loop plants. Wide variability was observed for the plant CDFs and for the CDFs of the contributing accident classes. On average, transients-with loss of injection, station blackout sequences, and transients with loss of decay heat removal are important contributors for the BWR 3/4 plants, while transients, station blackout sequences, and loss-of-coolant accidents are important for the Westinghouse 4-loop plants. The key factors that contribute to the variability in the results are discussed. The results are often driven by plant-specific design and operational characteristics, but differences in modeling approaches are also important for some accident classes

  17. Stability tests of the Westinghouse coil in the International Fusion Superconducting Magnet Test Facility

    International Nuclear Information System (INIS)

    Dresner, L.; Fehling, D.T.; Lubell, M.S.; Lue, J.W.; Luton, J.N.; McManamy, T.J.; Shen, S.S.; Wilson, C.T.

    1987-09-01

    The Westinghouse coil is one of three forced-flow coils in the six-coil toroidal array of the International Fusion Superconducting Magnet Test Facility at Oak Ridge National Laboratory. It is wound with an 18-kA, Nb 3 Sn/Cu, cable-in-conduit superconductor structurally supported by aluminum plates and cooled by 4-K, 15-atm supercritical helium. The coil is instrumented to permit measurement of helium temperature, pressure, and flow rate; structure temperature and strain; field; and normal zone voltage. A resistive heater has been installed to simulate nuclear heating, and inductive heaters have been installed to facilitate stability testing. The coil has been tested both individually and in the six-coil array. The tests covered charging to full design current and field, measuring the current-sharing threshold temperature using the resistive heaters, and measuring the stability margin using the pulsed inductive heaters. At least one section of the conductor exhibits a very broad resistive transition (resistive transition index = 4). The broad transition, though causing the appearance of voltage at relatively low temperatures, does not compromise the stability margin of the coil, which was greater than 1.1 J/cm 3 of strands. In another, nonresistive location, the stability margin was between 1.7 and 1.9 J/cm 3 of strands. The coil is completely stable in operation at 100% design current in both the single- and six-coil modes

  18. Station Blackout Analysis for a 3-Loop Westinghouse PWR Reactor Using Trace

    International Nuclear Information System (INIS)

    El-Sahlamy, N.M.

    2017-01-01

    One of the main concerns in the area of severe accidents in nuclear reactors is that of station blackout (SBO). The loss of offsite electrical power concurrent with the unavailability of the onsite emergency alternating current (AC) power system can result in loss of decay heat removal capability, leading to a potential core damage which may lead to undesirable consequences to the public and the environment. To cope with an SBO, nuclear reactors are provided with protection systems that automatically shut down the reactor, and with safety systems to remove the core residual heat. This paper provides a best estimate assessment of the SBO scenario in a 3-loop Westinghouse PWR reactor. The evaluation is performed using TRACE, a best estimate computer code for thermal-hydraulic calculations. Two sets of scenarios for SBO analyses are discussed in the current work. The first scenario is the short term SBO where it is assumed that in addition to the loss of AC power, there is no DC power; i.e., no batteries are available. In the second scenario, a long term SBO is considered. For this scenario, DC batteries are available for four hours. The aim of the current SBO analyses for the 3-loop pressurized water reactor presented in this paper is to focus on the effect of the availability of a DC power source to delay the time to core uncovers and heatup

  19. Westinghouse integrated protection system. An overview of the software design and maintenance features

    International Nuclear Information System (INIS)

    Gibson, R.J.

    1995-01-01

    The Westinghouse Integrated Protection System was designed with the goal of providing a system which can be easily verified, validated, and maintained. The software design and structure promote the ease of translation from functional requirements to applications function software while also improving the ability to verify and maintain the applications function software. The use of independent, reusable, common functions software modules focuses the design, verification, and validation of the software and reduces the likelihood of errors occurring during the application and maintenance of the software. The simple continuous loop method of operation used throughout the IPS provides a standard deterministic method of operation. The IPS design also incorporates the use of embedded self-diagnostics to perform continuous hardware oriented tests of the system and the use of an independent subsystem to automatically perform a functional test of the system. Maintenance interfaces also exist to readily identify and locate faults as well as providing other maintenance capabilities. These testing and maintenance features enhance the overall reliability and availability of the system. (orig.) (2 refs., 2 figs.)

  20. Signal validation of SPDS variables for Westinghouse and Combustion Engineering plants - an EPRI project

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    Signal validation in the context of this project is the process of combining information from multiple plant sensors to produce highly reliable information about plant conditions. High information reliability is achieved by the use of redundant sources of information and by the inherent detection, identification, and isolation of faulty signals. The signal validation methodology that has been developed in previous EPRI-sponsored projects has been enhanced and applied toward validation of critical safety-related SPDS signals in the Northeast Utilities Millstone 3 Westinghouse PWR plant and the Millstone 2 Combustion Engineering PWR plant. The designs were implemented in FORTRAN software and tested off-line using recorded plant sensor data, RETRAN-generated simulation data, and data to exercise software logic branches and the integration of software modules. Designs and software modules have been developed for 15 variables to support six PWR SPDS critical safety functions as required by a utility advisory group attached to the project. The signal validation process automates a task currently performed by plant operators and does so with consistent, verified logic regardless of operator stress and training level. The methodology uses a simple structure of generic software blocks, a modular implementation, and it performs effectively within the processor and memory constraints of modern plant process computers. The ability to detect and isolate sensor failures with greater sensitivity, robustness, and coverage of common-cause failures should ultimately lead to improved plant availability, efficiency, and productivity

  1. Westinghouse experience in Kozloduy NPP units 5 and 6 I and C modernization

    International Nuclear Information System (INIS)

    Sechensky, B.

    2005-01-01

    The paper presents the background, current implementation approach and experience on the largest ever modernization program on operating units WWER 1000 (PWR) at Kozloduy Nuclear Power Plant in Bulgaria. The Modernization Program itself includes more than 212 measures. Westinghouse is modernizing the major I and C systems at WWER 1000. The major topics of the modernization program and specific approach described in this paper are as follows: 1) Design Approach and Feature; 2) Installation Approach; 3) Test Strategy; 4) Licensing Strategy, applicable codes and standards. At the end author summarized that: 1) Specific design solutions were required and developed in order to address the specific plant features. At each stage, representatives of the Client are being involved in the process of designing and testing of the equipment and systems; 2) Phase-by-phase installation efforts were developed and extensive installation design documentation was prepared to fit in the limited outage window and to successfully complete the installation activities; 3) Well-prepared, multi-phase testing strategy was developed and is being implemented to assure the proper and adequate operation of the equipment at the factory and at the real plant

  2. Westinghouse Modular Grinding Process - Enhancement of Volume Reduction for Hot Resin Supercompaction - 13491

    Energy Technology Data Exchange (ETDEWEB)

    Fehrmann, Henning [Westinghouse Electric Germany GmbH, Dudenstr. 44, D-68167 Mannheim (Germany); Aign, Joerg [Westinghouse Electric Germany GmbH, Global D and D and Waste Management, Tarpenring 6, D-22419 Hamburg (Germany)

    2013-07-01

    In nuclear power plants (NPP) ion exchange (IX) resins are used in several systems for water treatment. Spent resins can contain a significant amount of contaminates which makes treatment for disposal of spent resins mandatory. Several treatment processes are available such as direct immobilization with technologies like cementation, bitumisation, polymer solidification or usage of a high integrity container (HIC). These technologies usually come with a significant increase in final waste volume. The Hot Resin Supercompaction (HRSC) is a thermal treatment process which reduces the resin waste volume significantly. For a mixture of powdered and bead resins the HRSC process has demonstrated a volume reduction of up to 75 % [1]. For bead resins only the HRSC process is challenging because the bead resins compaction properties are unfavorable. The bead resin material does not form a solid block after compaction and shows a high spring back effect. The volume reduction of bead resins is not as good as for the mixture described in [1]. The compaction properties of bead resin waste can be significantly improved by grinding the beads to powder. The grinding also eliminates the need for a powder additive.Westinghouse has developed a modular grinding process to grind the bead resin to powder. The developed process requires no circulation of resins and enables a selective adjustment of particle size and distribution to achieve optimal results in the HRSC or in any other following process. A special grinding tool setup is use to minimize maintenance and radiation exposure to personnel. (authors)

  3. Westinghouse-GOTHIC modeling of NUPEC's hydrogen mixing and distribution test M-4-3

    International Nuclear Information System (INIS)

    Ofstun, R.P.; Woodcock, J.; Paulsen, D.L.

    1994-01-01

    NUPEC (NUclear Power Engineering Corporation) ran a series of hydrogen mixing and distribution tests which were completed in April 1992. These tests were performed in a 1/4 linearly scaled model containment and were specifically designed to be used for computer code validation. The results of test M-4-3, along with predictions from several computer codes, were presented to the participants of ISP-35 (a blind test comparison of code calculated results with data from NUPEC test M-7-1) at a meeting in March 1993. Test M-4-3, which was similar to test M-7-1, released a mixture of steam and helium into a steam generator compartment located on the lower level of containment. The majority of codes did well at predicting the global pressure and temperature trends, however, some typical lumped parameter modeling problems were identified at that time. In particular, the models had difficulty predicting the temperature and helium concentrations in the so called 'dead ended volumes' (pressurizer compartment and in-core chase region). Modeling of the dead-ended compartments using a single lumped parameter volume did not yield the appropriate temperature and helium response within that volume. The Westinghouse-GOTHIC (WGOTHIC) computer code is capable of modeling in one, two or three dimensions (or any combination thereof). This paper describes the WGOTHIC modeling of the dead-ended compartments for NUPEC test M-4-3 and gives comparisons to the test data. 1 ref., 1 tab., 14 figs

  4. Recent improvements and new features in the Westinghouse lattice physics codes

    International Nuclear Information System (INIS)

    Huria, H.C.; Buechel, R.J.

    1995-01-01

    Westinghouse has been using the ANC three-dimensional, two-energy-group nodal model for nuclear analysis and fuel management calculations for standard pressurized water reactor (PWR) reload design analysis since 1988. The cross sections are obtained from PHOENIX-P, a modified version of the PHOENIX lattice physics code for all square-assembly PWR cores. The PHOENIX-H code was developed for modeling both the VVER-1000 and VVER-440 fuel lattice configurations. The PHOENIX-H code has evolved from PHOENIX-P, the primary difference being in the neutronic solution modules. The PHOENIX-P code determines the assembly flux distribution using integral transport theory-based pin-cell nodal coupling followed by two-dimensional discrete ordinates solution in x-y geometry. The PHOENIX-H code uses the two-dimensional heterogeneous response method. The other infrastructure is identical in both the codes, and they share the same 42-group cross-section library

  5. Los Alamos MAWST software layered on Westinghouse Savannah River Company's nuclear materials accountability system

    International Nuclear Information System (INIS)

    Whitty, W.J.; Smith, J.E.; Davis, J.M. Jr.

    1995-01-01

    The Los Alamos Safeguards Systems Group's Materials Accounting With Sequential Testing (MAWST) computer program was developed to fulfill DOE Order 5633.3B requiring that inventory-difference control limits be based on variance propagation or any other statistically valid technique. Westinghouse Savannah River Company (WSRC) developed a generic computerized accountability system, NucMAS, to satisfy accounting and reporting requirements for material balance areas. NucMAS maintains the calculation methods and the measurement information required to compute nuclear material transactions in elemental and isotopic masses by material type code. The Safeguards Systems Group designed and implemented to WSRC's specifications a software interface application, called NucMASloe. It is a layered product for NucMAS that automatically formats a NucMAS data set to a format compatible with MAWST and runs MAWST. This paper traces the development of NucMASloe from the Software Requirements through the testing and demonstration stages. The general design constraints are described as well as the difficulties encountered on interfacing an external software product (MAWST) with an existing classical accounting structure (NucMAS). The lessons learned from this effort, the design, and some of the software are directly applicable to the Local Area Network Material Accountability System (LANMAS) being sponsored by DOE

  6. Ichthyoplankton entrainment study at the SRS Savannah River water intakes for Westinghouse Savannah River Company

    International Nuclear Information System (INIS)

    Paller, M.

    1992-01-01

    Cooling water for L and K Reactors and makeup water for Par Pond is pumped from the Savannah River at the 1G, 3G, and 5G pump houses. Ichthyoplankton (drifting fish larvae and eggs) from the river are entrained into the reactor cooling systems with the river water and passed through the reactor's heat exchangers where temperatures may reach 70 degrees C during full power operation. Ichthyoplankton mortality under such conditions is assumed to be 100 percent. The number of ichthyoplankton entrained into the cooling system depends on a variety of variables, including time of year, density and distribution of ichthyoplankton in the river, discharge levels in the river, and the volume of water withdrawn by the pumps. Entrainment at the 1 G pump house, which is immediately downstream from the confluence of Upper Three Runs Creek and the Savannah River, is also influenced by discharge rates and ichthyoplankton densities in Upper Three Runs Creek. Because of the anticipated restart of several SRS reactors and the growing concern surrounding striped bass and American shad stocks in the Savannah River, the Department of Energy requested that the Environmental Sciences Section (ESS) of the Savannah River Laboratory sample ichthyoplankton at the SRS Savannah River intakes. Dams ampersand Moore, Inc., under a contract with Westinghouse Savannah River Company performed the sampling and data analysis for the ESS

  7. 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

  8. Aging assessment of Westinghouse PWR and General Electric BWR containment isolation functions

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B.S.; Travis, R.; Grove, E.; DiBiasio, A.

    1996-03-01

    A study was performed to assess the effects of aging on the Containment Isolation (CI) functions of Westinghouse Pressurized Water Reactors and General Electric Boiling Water Reactors. This study is part of the Nuclear Plant Aging Research (NPAR) program, sponsored by the U.S. Nuclear Regulatory Commission. The objectives of this program are to provide an understanding of the aging process and how it affects plant safety so that it can be properly managed. This is one of a number of studies performed under the NPAR program which provide a technical basis for the identification and evaluation of degradation caused by age. Failure data from two national databases, Nuclear Plant Reliability Data System (NPRDS) and Licensee Event Reports (LERs), as well as plant specific data were reviewed and analyzed to understand the effects of aging on the CI functions. This study provided information on the effects of aging on component failure frequency, failure modes, and failure causes. Current inspection, surveillance, and monitoring practices were also reviewed.

  9. Aging assessment of Westinghouse PWR and General Electric BWR containment isolation functions

    International Nuclear Information System (INIS)

    Lee, B.S.; Travis, R.; Grove, E.; DiBiasio, A.

    1996-03-01

    A study was performed to assess the effects of aging on the Containment Isolation (CI) functions of Westinghouse Pressurized Water Reactors and General Electric Boiling Water Reactors. This study is part of the Nuclear Plant Aging Research (NPAR) program, sponsored by the U.S. Nuclear Regulatory Commission. The objectives of this program are to provide an understanding of the aging process and how it affects plant safety so that it can be properly managed. This is one of a number of studies performed under the NPAR program which provide a technical basis for the identification and evaluation of degradation caused by age. Failure data from two national databases, Nuclear Plant Reliability Data System (NPRDS) and Licensee Event Reports (LERs), as well as plant specific data were reviewed and analyzed to understand the effects of aging on the CI functions. This study provided information on the effects of aging on component failure frequency, failure modes, and failure causes. Current inspection, surveillance, and monitoring practices were also reviewed

  10. Collections and Analyses of Common Cause Failure Data for the Korea Standard and Westinghouse Type NPPs

    International Nuclear Information System (INIS)

    Kang, Dae Il; Han, S. H.

    2007-04-01

    The analyses of the CCF events for domestic NPPs were performed to establish the domestic database for the CCF events and to deliver supply them to the operation office of the international common cause failure data exchange (ICDE) project. We collected and analyzed the CCF events of emergency diesel generators, centrifugal pumps, motor-operated valves, check valves, circuit breakers for the Korean Standard Type nuclear power plants (NPPs), Yonggwang Units 3 and 4 and Ulchin Units 3 and 4, and the Westinghouse type NPPs, Kori Unit 3 and 4 and Yonggwang Units 1 and 2. First, the components to be collected and analyzed were classified into the common cause component groups (CCCGs) according to the ICDE coding guidelines. Next, the CCF events were identified based on reviews of the component database for the PSA and its related documents, and consultations with NPP staff. Fourteen CCF events were identified. The ratio of the number of CCF events to that of individual failure events was identified as approximately 10 percentages. However, an in depth review of the CCF events showed that most failure severities of them were identified as partial CCF events, which can be interpreted as some component failures within the CCCGs. Root causes of the CCF events were identified as 9 internal part failures, 2 human errors, 2 design deficiencies, 1 procedure inadequacy. It could be concluded that the major root causes of the CCF events were internal piece part failures

  11. Evaluation of the rod ejection accident in Westinghouse Pressurized Water Reactors using spatial kinetics methods

    International Nuclear Information System (INIS)

    Risher, D.H. Jr.

    1975-01-01

    The consequences of a rod ejection accident are investigated in relation to the latest, high power density Westinghouse reactors. Limiting criteria are presented, based on experimental evidence, and if not exceeded these criteria will ensure that there will be no interference with core cooling capability, and radiation releases, if any, will be within the guidelines of 10CFR100. A basis is presented for the conservative selection of plant parameters to be used in the analysis, such that the analysis is applicable to a wide range of past, present, and future reactors. The calculational method employs a one-dimensional spatial kinetics computer code and a transient fuel heat transfer computer code to determine the hot spot fuel temperature versus time following a rod ejection. Using these computer codes, the most limiting hot channel factor (which does not cause the fuel damage limit criteria to be exceeded) has been determined as a function of the ejected rod worth. By this means, the limit criteria have been translated into ejected rod worths and hot channel factors which can be used effectively by the nuclear designer and safety analyst. The calculational method is shown to be conservative, compared to the results of a three-dimensional spatial kinetics analysis

  12. Identification of items and activities important to waste form acceptance by Westinghouse GoCo sites

    International Nuclear Information System (INIS)

    Plodinec, M.J.; Marra, S.L.; Dempster, J.; Randklev, E.H.

    1993-01-01

    The Department of Energy has established specifications (Waste Acceptance Product Specifications for Vitrified High-Level Waste Forms, or WAPS) for canistered waste forms produced at Hanford, Savannah River, and West Valley. Compliance with these specifications requires that each waste form producer identify the items and activities which must be controlled to ensure compliance. As part of quality assurance oversight activities, reviewers have tried to compare the methodologies used by the waste form producers to identify items and activities important to waste form acceptance. Due to the lack of a documented comparison of the methods used by each producer, confusion has resulted over whether the methods being used are consistent. This confusion has been exacerbated by different systems of nomenclature used by each producer, and the different stages of development of each project. The waste form producers have met three times in the last two years, most recently on June 28, 1993, to exchange information on each producer's program. These meetings have been sponsored by the Westinghouse GoCo HLW Vitrification Committee. This document is the result of this most recent exchange. It fills the need for a documented comparison of the methodologies used to identify items and activities important to waste form acceptance. In this document, the methodology being used by each waste form producer is summarized, and the degree of consistency among the waste form producers is determined

  13. Calculation of particulate dispersion in a design-basis tornadic storm from Westinghouse PFDL, Cheswick, Pennsylvania

    International Nuclear Information System (INIS)

    Pepper, D.W.

    1978-07-01

    A three-dimensional numerical model is used to calculate ground-level air concentration and deposition (due to precipitation scavenging) after a hypothetical tornado strike at the Westinghouse Plutonium Fuel Development Laboratory (PFDL) at Cheswick, Pennsylvania. Plutonium particles less than 20 μm in diameter are assumed to be lifted into the tornadic storm cell by the vortex. The rotational characteristics of the tornadic storm are embedded within the larger mesoscale flow of the storm system. The design-basis translational wind values are based on probabilities associated with existing records of tornado strikes in the vicinity of the plant site. Turbulence exchange coefficients are based on empirical values deduced from experimental data in severe storms and from theoretical assumptions obtained from the literature. The method of moments is used to incorporate subgrid-scale resolution of the concentration within a grid cell volume. This method is a quasi-Lagrangian scheme which minimizes numerical error associated with advection. In all case studies, the effects of updrafts and downdrafts, coupled with scavenging of the particulates by precipitation, account for most of the material being deposited within 20-45 km downwind of the plant site. Ground-level isopleths in the x-y plane show that most of the material is deposited behind and slightly to the left of the centerline trajectory of the storm. Approximately 5% of the material is dispersed into the stratosphere and anvil section of the storm

  14. Quantification of severe accident source terms of a Westinghouse 3-loop plant

    International Nuclear Information System (INIS)

    Lee Min; Ko, Y.-C.

    2008-01-01

    Integrated severe accident analysis codes are used to quantify the source terms of the representative sequences identified in PSA study. The characteristics of these source terms depend on the detail design of the plant and the accident scenario. A historical perspective of radioactive source term is provided. The grouping of radionuclides in different source terms or source term quantification tools based on TID-14844, NUREG-1465, and WASH-1400 is compared. The radionuclides release phenomena and models adopted in the integrated severe accident analysis codes of STCP and MAAP4 are described. In the present study, the severe accident source terms for risk quantification of Maanshan Nuclear Power Plant of Taiwan Power Company are quantified using MAAP 4.0.4 code. A methodology is developed to quantify the source terms of each source term category (STC) identified in the Level II PSA analysis of the plant. The characteristics of source terms obtained are compared with other source terms. The plant analyzed employs a Westinghouse designed 3-loop pressurized water reactor (PWR) with large dry containment

  15. Westinghouse Hanford Company plan for certifying newly generated contact -- handled transuranic waste. Revision 1

    International Nuclear Information System (INIS)

    Lipinski, R.M.; Backlund, E.G.

    1995-09-01

    All transuranic (TRU) waste generators are required by US Department of Energy (DOE) Order 5820.2A to package their TRU waste in order to comply wit the Waste Isolation Pilot Plant (WIPP) -- Waste Acceptance Criteria (WAC) or keep non-certifiable containers segregated. The Westinghouse Hanford Company (WHC) Transuranic Waste Certification Plan was developed to ensure that TRU newly generated waste at WHC meets the DOE Order 5820.2A and the WHC-WAC which includes the State of Washington Department of Ecology -- Washington Administrative Code (DOE-WAC). The metho used at WHC to package TRU waste are described in sufficient detail to meet the regulations. This document is organized to provide a brief overview of waste generation operations at WHC. The methods used to implement this plan are discussed briefly along with the responsibilities and authorities of applicable organizations. This plan describes how WHC complies with all applicable regulations and requirements set forth in the latest approved revision of WHC-EP-0063-4

  16. Assessment of the TASS 1-D neutronics model for the westinghouse and ABB-CE type PWR reactivity induced transients

    International Nuclear Information System (INIS)

    Choi, J.D.; Yoon, H.Y.; Um, K.S.; Kim, H.C.; Sim, S.K.

    1997-01-01

    Best estimate transient analysis code, TASS, has been developed for the normal and transient simulation of the Westinghouse and ABB-CE type PWRs. TASS thermal hydraulic model is based on the non-homogeneous, non-equilibrium two-phase continuity, energy and mixture momentum equations with constitutive relations for closure. Core neutronics model employs both the point kinetics and one-dimensional neutron diffusion model. Semi-implicit numerical scheme is used to solve the discretized finite difference equations. TASS one dimensional neutronics core model has been assessed through the reactivity induced transient analyses for the KORI-3, three loop Westinghouse PWR, and Younggwang-3 (YGN-3), two-loop ABB-CE PWR, nuclear power plants currently operating in Korea. The assessment showed that the TASS one dimensional neutronics core model can be applied for the Westinghouse and ABB-CE type PWRs to gain thermal margin which is necessary for a potential use of the high fuel burnup, extended fuel cycle, power upgrading and for the plant life extension

  17. Notes on the Implementation of Non-Parametric Statistics within the Westinghouse Realistic Large Break LOCA Evaluation Model (ASTRUM)

    International Nuclear Information System (INIS)

    Frepoli, Cesare; Oriani, Luca

    2006-01-01

    In recent years, non-parametric or order statistics methods have been widely used to assess the impact of the uncertainties within Best-Estimate LOCA evaluation models. The bounding of the uncertainties is achieved with a direct Monte Carlo sampling of the uncertainty attributes, with the minimum trial number selected to 'stabilize' the estimation of the critical output values (peak cladding temperature (PCT), local maximum oxidation (LMO), and core-wide oxidation (CWO A non-parametric order statistics uncertainty analysis was recently implemented within the Westinghouse Realistic Large Break LOCA evaluation model, also referred to as 'Automated Statistical Treatment of Uncertainty Method' (ASTRUM). The implementation or interpretation of order statistics in safety analysis is not fully consistent within the industry. This has led to an extensive public debate among regulators and researchers which can be found in the open literature. The USNRC-approved Westinghouse method follows a rigorous implementation of the order statistics theory, which leads to the execution of 124 simulations within a Large Break LOCA analysis. This is a solid approach which guarantees that a bounding value (at 95% probability) of the 95 th percentile for each of the three 10 CFR 50.46 ECCS design acceptance criteria (PCT, LMO and CWO) is obtained. The objective of this paper is to provide additional insights on the ASTRUM statistical approach, with a more in-depth analysis of pros and cons of the order statistics and of the Westinghouse approach in the implementation of this statistical methodology. (authors)

  18. Definition of thermal-hydraulics parameters of a naval PWR via energy balance of a Westinghouse PWR

    Energy Technology Data Exchange (ETDEWEB)

    Chaves, Luiz C.; Curi, Marcos F., E-mail: marcos.curi@cefet-rj.br [Centro Federal de Educação Tecnológica Celso Suckow da Fonseca (CEFET-RJ), Rio de Janeiro, RJ (Brazil). Department of Mechanical Engineering

    2017-07-01

    In this work, we used the operational parameters of the Angra 1 nuclear power plant, designed by Westinghouse, to estimate the thermal-hydraulic parameters for naval nuclear propulsion, focusing on the analysis of the reactor and steam generator. A thermodynamics analysis was made to reach the operational parameters of primary circuit such as pressure, temperature, power generated among others. Previous studies available in literature of 2-loop Westinghouse Nuclear Power Plants, which is based on a PWR and similar to Angra-1, support this analysis in the sense of a correct procedure to deal with many complex processes to energy generation from a nuclear source. Temperature profiles in reactor and steam generator were studied with concepts of heat transfer, fluid mechanics and also some concepts of nuclear systems, showing the behavior into them. In this simulation, the Angra 1 primary circuit was reduced on a scale of 1: 3.5 to fit in a Scorpène-class submarine. The reactor generates 85.7 MW of total thermal power. The maximum power and temperatures reached were lower than the operational safe limits established by Westinghouse. The number of tubes of the steam generator was determined in 990 U-tubes with 6.3 m of average length. (author)

  19. Westinghouse Hanford Company plan for certifying newly generated contact-handled transuranic waste for emplacement in the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Lipinski, R.M.; Sheehan, J.S.

    1992-07-01

    Westinghouse Hanford Company (Westinghouse Hanford) currently manages an interim storage site for Westinghouse Hanford and non-Westinghouse Hanford-generated transuranic (TRU) waste and operates TRU waste generating facilities within the Hanford Site in Washington State. Approval has been received from the Waste Acceptance Criteria Certification Committee (WACCC) and Westinghouse Hanford TRU waste generating facilities to certify newly generated contact-handled TRU (CH-TRU) solid waste to meet the Waste Acceptance Criteria (WAC). This document describes the plan for certifying newly generated CH-TRU solid waste to meet the WAC requirements for storage at the Waste Isolation Pilot Plant (WIPP) site. Attached to this document are facility-specific certification plans for the Westinghouse Hanford TRU waste generators that have received WACCC approval. The certification plans describe operations that generate CH-TRU solid waste and the specific procedures by which these wastes will be certified and segregated from uncertified wastes at the generating facilities. All newly generated CH-TRU solid waste is being transferred to the Transuranic Storage and Assay Facility (TRUSAF) and/or a controlled storage facility. These facilities will store the waste until the certified TRU waste can be sent to the WIPP site and the non-certified TRU waste can be sent to the Waste Receiving and Processing Facility. All non-certifiable TRU waste will be segregated and clearly identified

  20. Spanish program of advanced Nuclear Power Plants

    International Nuclear Information System (INIS)

    Marco, M.; Redon, R.

    1993-01-01

    The energy Spanish Plan is promoting some actions within the area of advanced reactors. Efforts are focussed onto the European Program of Advanced Reactors, the Program of Passive Plants (EPRI), European Fast Reactor Project and the APWR-1000 Program of INI. Electrical sector utilities and industrial partners supported by the Administration have organized an steering committee. The program of Passive Plants includes activities on Qualification, design and detailed engineering (Qualification project, SBWR project of G.E. and AP600 Project of Westinghouse. The european project on advanced plants has the following Spanish contribution: Analysis of alternative Dossier on European requisites (EUR) and Design of an European Reactor (EPR)

  1. Application of CASMO-4/MICROBURN-B2 methodology to mixed cores with Westinghouse Optima2 fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hsiao, Ming Yuan; Wheeler, John K.; Hoz, Carlos de la [Nuclear Fuels, Warrenville (United States)

    2008-10-15

    The first application of CASMO-4/MICROBURN-B2 methodology to Westinghouse SVEA-96 Optima2 reload cycle is described in this paper. The first Westinghouse Optima2 reload cycle in the U.S. is Exelon's Quad Cities Unit 2 Cycle 19 (Q2C19). The core contains fresh Optima2 fuel and once burned and twice burned GE14 fuel. Although the licensing analyses for the reload cycle are performed by Westinghouse with Westinghouse methodology, the core is monitored with AREVA's POWERPLEX-III core monitoring system that is based on the CASMO-4/MICROBURN-B2 (C4/B2) methodology. This necessitates the development of a core model based on the C4/B2 methodology for both reload design and operational support purposes. In addition, as expected, there are many differences between the two vendors' methodologies; they differ not only in modeling some of the physical details of the Optima2 bundles but also in the modeling capability of the computer codes. In order to have high confidence that the online core monitoring results during the cycle startup and operation will comply with the Technical Specifications requirements (e.g., thermal limits, shutdown margins), the reload core design generated by Westinghouse design methodology was confirmed by the C4/B2 model. The C4/B2 model also assures that timely operational support during the cycle can be provided. Since this is the first application of C4/B2 methodology to an Optima2 reload in the US, many issues in the lattice design, bundle design, and reload core design phases were encountered. Many modeling issues have to be considered in order to develop a successful C4/B2 core model for the Optima2/GE14 mixed core. Some of the modeling details and concerns and their resolutions are described. The Q2C19 design was successfully completed and the 2 year cycle successfully started up in April 2006 and shut down in March 2008. Some of the operating results are also presented.

  2. Application of CASMO-4/MICROBURN-B2 methodology to mixed cores with Westinghouse Optima2 fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hsiao, Ming Yuan; Wheeler, John K.; Hoz, Carlos de la [Nuclear Fuels, Warrenville (United States)

    2008-10-15

    The first application of CASMO-4/MICROBURN-B2 methodology to Westinghouse SVEA-96 Optima2 reload cycle is described in this paper. The first Westinghouse Optima2 reload cycle in the U.S. is Exelon's Quad Cities Unit 2 Cycle 19 (Q2C19). The core contains fresh Optima2 fuel and once burned and twice burned GE14 fuel. Although the licensing analyses for the reload cycle are performed by Westinghouse with Westinghouse methodology, the core is monitored with AREVA's POWERPLEX-III core monitoring system that is based on the CASMO-4/MICROBURN-B2 (C4/B2) methodology. This necessitates the development of a core model based on the C4/B2 methodology for both reload design and operational support purposes. In addition, as expected, there are many differences between the two vendors' methodologies; they differ not only in modeling some of the physical details of the Optima2 bundles but also in the modeling capability of the computer codes. In order to have high confidence that the online core monitoring results during the cycle startup and operation will comply with the Technical Specifications requirements (e.g., thermal limits, shutdown margins), the reload core design generated by Westinghouse design methodology was confirmed by the C4/B2 model. The C4/B2 model also assures that timely operational support during the cycle can be provided. Since this is the first application of C4/B2 methodology to an Optima2 reload in the US, many issues in the lattice design, bundle design, and reload core design phases were encountered. Many modeling issues have to be considered in order to develop a successful C4/B2 core model for the Optima2/GE14 mixed core. Some of the modeling details and concerns and their resolutions are described. The Q2C19 design was successfully completed and the 2 year cycle successfully started up in April 2006 and shut down in March 2008. Some of the operating results are also presented.

  3. Application of CASMO-4/MICROBURN-B2 methodology to mixed cores with Westinghouse Optima2 fuel

    International Nuclear Information System (INIS)

    Hsiao, Ming Yuan; Wheeler, John K.; Hoz, Carlos de la

    2008-01-01

    The first application of CASMO-4/MICROBURN-B2 methodology to Westinghouse SVEA-96 Optima2 reload cycle is described in this paper. The first Westinghouse Optima2 reload cycle in the U.S. is Exelon's Quad Cities Unit 2 Cycle 19 (Q2C19). The core contains fresh Optima2 fuel and once burned and twice burned GE14 fuel. Although the licensing analyses for the reload cycle are performed by Westinghouse with Westinghouse methodology, the core is monitored with AREVA's POWERPLEX-III core monitoring system that is based on the CASMO-4/MICROBURN-B2 (C4/B2) methodology. This necessitates the development of a core model based on the C4/B2 methodology for both reload design and operational support purposes. In addition, as expected, there are many differences between the two vendors' methodologies; they differ not only in modeling some of the physical details of the Optima2 bundles but also in the modeling capability of the computer codes. In order to have high confidence that the online core monitoring results during the cycle startup and operation will comply with the Technical Specifications requirements (e.g., thermal limits, shutdown margins), the reload core design generated by Westinghouse design methodology was confirmed by the C4/B2 model. The C4/B2 model also assures that timely operational support during the cycle can be provided. Since this is the first application of C4/B2 methodology to an Optima2 reload in the US, many issues in the lattice design, bundle design, and reload core design phases were encountered. Many modeling issues have to be considered in order to develop a successful C4/B2 core model for the Optima2/GE14 mixed core. Some of the modeling details and concerns and their resolutions are described. The Q2C19 design was successfully completed and the 2 year cycle successfully started up in April 2006 and shut down in March 2008. Some of the operating results are also presented

  4. Westinghouse Savannah River Site vendor forum: An innovative cooperative technology development success

    International Nuclear Information System (INIS)

    Sturm, H.F. Jr.

    1996-01-01

    The Westinghouse Savannah River Company (WSRC) Supplier Environmental and Waste Management Information Exchange Forum was held August 31 - September 1, 1993. The forum, which was planned and conducted in concert with the Department of Energy Savannah River Operations Office (DOE-SROO), was held to foster a technical exchange in which new, innovative technologies were proposed by suppliers, to identify more cost-effective methods to apply to future and on-going activities, to increase use of the private sector, and to promote partnerships with other industries. The two day forum provided the opportunity for WSRC and DOE-SR to review program activities and challenges in five major areas, Savannah River Technology Center, Solid Waste Facilities, Environmental Restoration, Environmental Monitoring, and Decontamination and Decommissioning through formal presentations. The second day was designed to provide suppliers the opportunity to talk about current and future activities and challenges with representatives of each of these areas at display booths, special high interest topic interactive sessions, and site tours. Each attendee was then invited to submit pre-proposals relative to the abstracts presented in The Special Consolidate Solicitation for Environmental and Waste Management Basic and Applied Research and Research-Related Development and/or Demonstration No. E10600-E1 document. Twenty-five contracts totaling $12 million were awarded. Twenty-four contracts have now been completed. This paper provides an overview of the pre forum activities, the forum, post-forum and proposal review process, and most importantly a description of the technologies demonstrated, the benefits and savings derived, and future use potential from a DOE perspective, as well as technology transfer and industrial partnership potential

  5. Lessons Learned From Implementation of Westinghouse Owners Group Risk-Informed Inservice Inspection Methodology for Piping

    International Nuclear Information System (INIS)

    Stevenson, Paul R.; Haessler, Richard L.; McNeill, Alex; Pyne, Mark A.; West, Raymond A.

    2006-01-01

    Risk-informed inservice inspection (ISI) programs have been in use for over seven years as an alternative to current regulatory requirements in the development and implementation of ISI programs for nuclear plant piping systems. Programs using the Westinghouse Owners Group (WOG) (now known as the Pressurized Water Reactor Owners Group - PWROG) risk-informed ISI methodology have been developed and implemented within the U.S. and several other countries. Additionally, many plants have conducted or are in the process of conducting updates to their risk-informed ISI programs. In the development and implementation of these risk-informed ISI programs and the associated updates to those programs, the following important lessons learned have been identified and are addressed. Concepts such as 'loss of inventory', which are typically not modeled in a plant's probabilistic risk assessment (PRA) model for all systems. The importance of considering operator actions in the identification of consequences associated with a piping failure and the categorization of segments as high safety significant (HSS) or low safety significant (LSS). The impact that the above considerations have had on the large early release frequency (LERF) and categorization of segments as HSS or LSS. The importance of automation. Making the update process more efficient to reduce costs associated with maintaining the risk-informed ISI program. The insights gained are associated with many of the steps in the risk-informed ISI process including: development of the consequences associated with piping failures, categorization of segments, structural element selection and program updates. Many of these lessons learned have impacted the results of the risk-informed ISI programs and have impacted the updates to those programs. This paper summarizes the lessons learned and insights gained from the application of the WOG risk-informed ISI methodology in the U.S., Europe and Asia. (authors)

  6. The AP600 advanced simplified nuclear power plant. Results of the test program and progress made toward final design approval

    International Nuclear Information System (INIS)

    Bruschi, H.J.

    1996-01-01

    At the 1994 Pacific Basin Conference, Mr. Bruschi presented a paper describing the AP600, Westinghouse's advanced light water reactor design with passive safety features. Since then, a rigorous test program was completed and AP600 became the most thoroughly tested advanced reactor system design in history. Westinghouse is now well on its way toward receiving Final Design Approval from the U.S. Nuclear Regulatory Commission for AP600. In this paper, the results of the test program will be discussed and an update on prospects for building the plant will be covered. (author)

  7. Numerical simulation of multi-phase phenomena in IVR related processes

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, Xu [Karlsruher Institut fuer Technologie (KIT), Karlsruhe (Germany). Bereich Innovative Reaktorsysteme; Shanghai Jiao Tong Univ. (China). School of Nuclear Science and Engineering

    2016-05-15

    IVR (in-vessel retention) is one of the severe accident mitigation measures, which is widely applied in the advanced light water reactors (LWRs) such as KERENA of AREVA, AP1000 of Westinghouse and CAP1400 of SNPTC, and attracts extensive interests of the German and Chinese nuclear scientists. The ultimate target of IVR is to keep the core melt inside the reactor pressure vessel (RPV) and to provide cooling capability via water flowing outside the RPV, the so called external reactor vessel cooling (ERVC). This paper summarizes some activities ongoing in both KIT and SJTU (Shanghai Jiao Tong University) with a few results examples.

  8. Stem thrust prediction model for Westinghouse wedge gate valves with linkage type stem-to-disk connection

    International Nuclear Information System (INIS)

    Wang, J.K.; Sharma, V.; Kalsi, M.S.

    1996-01-01

    The Electric Power Research Institute (EPRI) conducted a comprehensive research program with the objective of providing nuclear utilities with analytical methods to predict motor operated valve (MOV) performance under design basis conditions. This paper describes the stem thrust calculation model developed for evaluating the performance of one such valve, the Westinghouse flexible wedge gate valve. These procedures account for the unique functional characteristics of this valve design. In addition, model results are compared to available flow loop and in situ test data as a basis for evaluating the performance of the valve model

  9. Stem thrust prediction model for Westinghouse wedge gate valves with linkage type stem-to-disk connection

    Energy Technology Data Exchange (ETDEWEB)

    Wang, J.K.; Sharma, V.; Kalsi, M.S. [Kalsi Engineering, Inc., Sugar Land, TX (United States)] [and others

    1996-12-01

    The Electric Power Research Institute (EPRI) conducted a comprehensive research program with the objective of providing nuclear utilities with analytical methods to predict motor operated valve (MOV) performance under design basis conditions. This paper describes the stem thrust calculation model developed for evaluating the performance of one such valve, the Westinghouse flexible wedge gate valve. These procedures account for the unique functional characteristics of this valve design. In addition, model results are compared to available flow loop and in situ test data as a basis for evaluating the performance of the valve model.

  10. Westinghouse Hanford Company effluent discharges and solid waste management report for calendar year 1989: 200/600 Areas

    International Nuclear Information System (INIS)

    Brown, M.J.; P'Pool, R.K.; Thomas, S.P.

    1990-05-01

    This report presents calendar year 1989 radiological and nonradiological effluent discharge data from facilities in the 200 Areas and the 600 Area of the Hanford Site. Both summary and detailed effluent data are presented. In addition, radioactive and nonradioactive solid waste storage and disposal data for calendar year 1989 are furnished. Where appropriate, comparisons to previous years are made. The intent of the report is to demonstrate compliance of Westinghouse Hanford Company-operated facilities with administrative control values for radioactive constituents and applicable guidelines and standards (including Federal permit limits) for nonradioactive constituents. 11 refs., 20 tabs

  11. Westinghouse Reference Safety Analysis Report, RESAR-414. License application, preliminary safety analysis report (RESAR-414) volume 1

    International Nuclear Information System (INIS)

    1976-01-01

    Westinghouse's standardized four-loop, single unit NSSS for a pressurized water reactor is described including the core, coolant system, ECCS, emergency boration, chemical and volume control, RHR system, boron recycle, fuel handling, spent fuel pool and associated instrumentation and controls. This reactor is applicable to a plant with a core power level of 3800 MW(t) and 1295 MW(e). The reactor is controlled by temperature coefficients of reactivity; control rod motion, and by a soluble neutron absorber-boric acid

  12. Evaluation of the Westinghouse 10B depletion for BWR control rods

    International Nuclear Information System (INIS)

    Vallgren, Christina

    2008-03-01

    The aim of this work was to establish the 10 B depletion model for CR 99 control rods by using the latest version of POLCA7. In order to obtain an understanding of the differences between the currently used 10 B depletion models implemented in POLCA4 at O3 and in SIMULATE-3 at OL1, and the latest improved model implemented in the latest POLCA7, this work has been performed in three different parts. The first part of the work was to find out how large differences there exist in 10 B depletion between the calculated data by using the latest core monitoring system (POLCA7 version 4.10.0) and the measured data obtained in the hot-cell laboratory in Studsvik. It was found that the 10 B depletion computed by the latest POLCA7 version is in good agreement with the measured data from Studsvik. A poor agreement with a conservative overestimation in 10 B depletion was also found between the old model and the measured data. The aim of the second part of the work was to compare the calculated 10 B depletion values for two CR 99 rods from Olkiluoto 1 with the calculated 10 B depletion value for a CR 99 rod from Oskarshamn 3, by using the new 10 B depletion model implemented in the latest POLCA7 version. Swelling measurements of the boron carbide pins, used as absorber material, have indicated that the 10 B depletion should be of similar magnitude for the rods in Olkiluoto 1 and the rod in Oskarshamn 3, whereas the calculated values by using the earlier 10 B depletion models on the process computers showed a difference of about 20 %. By using the new 10 B depletion model m POLCA7, it was found that the 10 B depletion in the two studied cases was similar to each other and, thus, the hypothesis of a linear relationship between B 4 C swelling and thermal neutron fluence was supported. This third part of the work was carried out at KKL, Switzerland, and focused on comparing the B depletion models used in Westinghouse/POLCA7 and KKL/PRESTO-2. It was found that there is a slight

  13. Evaluation of the Westinghouse 10B depletion for BWR control rods

    Energy Technology Data Exchange (ETDEWEB)

    Vallgren, Christina

    2008-03-15

    The aim of this work was to establish the 10B depletion model for CR 99 control rods by using the latest version of POLCA7. In order to obtain an understanding of the differences between the currently used 10B depletion models implemented in POLCA4 at O3 and in SIMULATE-3 at OL1, and the latest improved model implemented in the latest POLCA7, this work has been performed in three different parts. The first part of the work was to find out how large differences there exist in 10B depletion between the calculated data by using the latest core monitoring system (POLCA7 version 4.10.0) and the measured data obtained in the hot-cell laboratory in Studsvik. It was found that the 10B depletion computed by the latest POLCA7 version is in good agreement with the measured data from Studsvik. A poor agreement with a conservative overestimation in 10B depletion was also found between the old model and the measured data. The aim of the second part of the work was to compare the calculated 10B depletion values for two CR 99 rods from Olkiluoto 1 with the calculated 10B depletion value for a CR 99 rod from Oskarshamn 3, by using the new 10B depletion model implemented in the latest POLCA7 version. Swelling measurements of the boron carbide pins, used as absorber material, have indicated that the 10B depletion should be of similar magnitude for the rods in Olkiluoto 1 and the rod in Oskarshamn 3, whereas the calculated values by using the earlier 10B depletion models on the process computers showed a difference of about 20 %. By using the new 10B depletion model m POLCA7, it was found that the 10B depletion in the two studied cases was similar to each other and, thus, the hypothesis of a linear relationship between B{sub 4}C swelling and thermal neutron fluence was supported. This third part of the work was carried out at KKL, Switzerland, and focused on comparing the B depletion models used in Westinghouse/POLCA7 and KKL/PRESTO-2. It was found that there is a slight difference in

  14. Prototyping Advanced Control Systems on FPGA

    Directory of Open Access Journals (Sweden)

    Simard Stéphane

    2009-01-01

    Full Text Available In advanced digital control and mechatronics, FPGA-based systems on a chip (SoCs promise to supplant older technologies, such as microcontrollers and DSPs. However, the tackling of FPGA technology by control specialists is complicated by the need for skilled hardware/software partitioning and design in order to match the performance requirements of more and more complex algorithms while minimizing cost. Currently, without adequate software support to provide a straightforward design flow, the amount of time and efforts required is prohibitive. In this paper, we discuss our choice, adaptation, and use of a rapid prototyping platform and design flow suitable for the design of on-chip motion controllers and other SoCs with a need for analog interfacing. The platform consists of a customized FPGA design for the Amirix AP1000 PCI FPGA board coupled with a multichannel analog I/O daughter card. The design flow uses Xilinx System Generator in Matlab/Simulink for system design and test, and Xilinx Platform Studio for SoC integration. This approach has been applied to the analysis, design, and hardware implementation of a vector controller for 3-phase AC induction motors. It also has contributed to the development of CMC's MEMS prototyping platform, now used by several Canadian laboratories.

  15. Plant maintenance and advanced reactors, 2006

    Energy Technology Data Exchange (ETDEWEB)

    Agnihotri, Newal (ed.)

    2006-09-15

    The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Advanced plants to meet rising expectations, by John Cleveland, International Atomic Energy Agency, Vienna; A flexible and economic small reactor, by Mario D. Carelli and Bojan Petrovic, Westinghouse Electric Company; A simple and passively safe reactor, by Yury N. Kuznetsov, Research and Development Institute of Power Engineering (NIKIET), Russia; Gas-cooled reactors, by Jeffrey S. Merrifield, U.S. Nuclear Regulatory Commission; ISI project managment in the PRC, by Chen Chanbing, RINPO, China; and, Fort Calhoun refurbishment, by Sudesh Cambhir, Omaha Public Power District.

  16. Advanced alarm management system

    International Nuclear Information System (INIS)

    Easter, J.R.

    1995-01-01

    The Westinghouse Advanced Alarm Management System (AWARE) is one of the Man-Machine Design Interfaces (MMI) which has great flexibility with regard to hardware type and configuration, alarm system concept, plant scope, engineering scope and installation. The AWARE System provides the capability to better manage the quantity prioritization and presentation of real-time process alarm messages in the control room. The messages are specific, precise and dynamic. The AWARE System can provide a large reduction in the number of messages that the control room staff must address at any one time, thus making the alarm message system a useful tool for the operators during situations that normally produce a high volume of messages as well as improving the clarity of the presentation of process abnormalities during small disturbances. The operating staff is now provided with the basis for a better understanding of the current plant state and for taking the appropriate control actions. (2 refs., 3 figs.)

  17. Licensing and advanced fuel designs

    International Nuclear Information System (INIS)

    Davidson, S.L.; Novendstern, E.H.

    1991-01-01

    For the past 15 years, Westinghouse has been actively involved in the development and licensing of fuel designs that contain major advanced features. These designs include the optimized fuel assembly, The VANTAGE 5 fuel assembly, the VANTAGE 5H, and most recently the VANTAGE+ fuel assembly. Each of these designs was supported by extensive experimental data, safety evaluations, and design efforts and required intensive interaction with the US Nuclear Regulatory Commission (NRC) during the review and approval process. This paper presents a description of the licensing approach and how it was utilized by the utilities to facilitate the licensing applications of the advanced fuel designs for their plants. The licensing approach described in this paper has been successfully applied to four major advanced fuel design changes ∼40 plant-specific applications, and >350 cycle-specific reloads in the past 15 years

  18. Quality assurance (QA) training at Westinghouse including innovative approaches for achieving an effective QA programme and establishing constructive interaction

    International Nuclear Information System (INIS)

    Chivers, J.H.; Scanga, B.E.

    1982-01-01

    Experience of the Westinghouse Water Reactors Division with indoctrination and training of quality engineers includes training of personnel from Westinghouse divisions in the USA and overseas as well as of customers' personnel. A written plan is prepared for each trainee in order to fit the training to the individual's needs, and to cover the full range of information and activities. The trainee is also given work assignments, working closely with experienced quality engineers. He may prepare inspection plans and audit check lists, assist in the preparation of QA training modules, write procedures, and perform supplier surveillance and data analyses, or make special studies of operating systems. The trainee attends seminars and special courses on work-related technical subjects. Throughout the training period, emphasis is placed on inculcating an attitude of team work in the trainee so that the result of the training is the achievement of both quality and productivity. Certification is extended (given that education/experience/skill requirements are met) to such functions as mechanical equipment quality engineering, electrical equipment quality engineering, and start-up and testing quality engineering. A well-trained quality engineer is equipped to provide technical assistance to other disciplines and, through effective co-operation with others, contributes to the success of the organization's endeavours. (author)

  19. Westinghouse Owners Group Risk-Informed Regulation Efforts: Options 2 and 3

    International Nuclear Information System (INIS)

    Brown, Jason A.; Osterrieder, Robert A.; Lutz, Robert J.; Dingler, Maurice; Ward, Lewis A.

    2002-01-01

    The U.S. Nuclear Regulatory Commission (NRC) has initiated efforts to incorporate risk-informed methods to redefine the scope of the existing 10 CFR 50 regulations (Option 2) and to change the technical requirements of the regulations (Option 3). The overall objectives of these efforts are to enhance plant safety, provide a framework for risk-informed regulations, add flexibility to plant operations, and reduce regulatory burden. The Westinghouse Owners Group (WOG) has a variety of active programs in the risk-informed area, including a program in the Option 2 and Option 3 areas. These two programs will be summarized including the benefits and the technical approach. The purpose of Option 2 is to make changes to the overall scope of structures, systems and components (SSCs) covered by 10 CFR 50 requiring special treatment by formulating new risk-informed safety classification categories that are linked to current definitions of safety-related and important-to-safety. This initiative would permit possible changes to the current special treatment requirements based on risk insights. The Nuclear Energy Institute (NEI) has developed an Option 2 implementation guideline (NEI 00-04 Draft Revision B). The WOG has initiated a program to validate the NEI guideline and to provide an initial cost-benefit assessment of the revised categorization and treatment under Option 2 via trial application to two systems at both Surry Unit 1 and Wolf Creek. The WOG Option 2 program includes consideration of all of the components in the selected systems, regardless of whether or not they are modeled in the respective plant probabilistic risk assessment (PRA) studies. As a result, quantitative risk measures are not available for many of the components being considered. In this case, the WOG program will provide valuable input to the NEI guideline. Additionally, the WOG program extends the use of both of the dominant methodologies for risk-informed ISI (RI-ISI) to address repair and

  20. Comparative economic analysis of the Integral Molten Salt Reactor and an advanced PWR using the G4-ECONS methodology

    International Nuclear Information System (INIS)

    Samalova, Ludmila; Chvala, Ondrej; Maldonado, G. Ivan

    2017-01-01

    The assessment of economic viability of a new reactor concept is crucial particularly during the early stages of its concept development. The G4-ECONS methodology provides a standardized top-down estimate of electricity cost and parametric sensitivities, not specifically targeted toward an accurate prediction of the final cost when deployed, but rather seeking an approximation of cost variations relative to other systems. This study presents an analysis of the Integral Molten Salt Reactor (IMSR) concept in comparison with a consistent analysis of an advanced PWR reactor (represented by AP1000). Estimation of levelized unit electricity costs, as well as sensitivity analyses to the discount rate and uranium or SWU prices, are presented using this methodology.

  1. Severe accident management development program for VVER-1000 and VVER-440/213 based on the westinghouse owners group approach

    International Nuclear Information System (INIS)

    Felix, E.; Dessars, N.

    2003-01-01

    The development of the Westinghouse Owners Group Severe Accident Management Guidelines (WOG SAMG) between 1991 and 1994 was initiated in response to the U.S. Nuclear Regulatory Commission (NRC) requirement for addressing the regulatory severe accident concerns. Hence, the WOG SAMG is designed to interface with other existing procedures at the plant and is used in accident sequences that have progressed to the point where these other procedures are not applicable any longer, i.e. following core damage. The primary purpose of the WOG SAMG is to reach a controlled stable state, which can be declared when fission product releases are controlled, challenges to the confinement fission product boundary have been mitigated, and adequate heat removal is provided to the core and the containment. Although the WOG SAMG is a generic severe accident management guidance developed for use by the entirety of the operating Westinghouse PWR plants, provisions have been made in their development to address specific features of individual plants such as confinement type and the feasibility of reactor cavity flooding. Similarly, the generic SAMG does not address unique plant features and equipment, but rather allows for consideration of plant specific features and strategies. This adaptable approach has led to several SAMG development programs for VVER-1000 and VVER-440 type of power plants, under Westinghouse' s lead. The first of these programs carried out to completion was for Temelin NPP - VVER-1000 - in the first quarter of 2003. Other ongoing programs aim at providing a similar work for VVER-440 design, namely Dukovany, Mochovce and Bohunice NPPs. The challenge of adapting the existing generic WOG material to plants other than PWRs mainly arises for VVER-440 because of important differences in confinement design, making it more vulnerable to ex-vessel phenomena such as hydrogen burn. Also, for both eastern designs, cavity flooding strategy requires special consideration and

  2. In the matter of the application of the Westinghouse Electric Corporation for the export of pressurized water reactor to Asociacion Nuclear ASCO II, Barcelona, Spain

    International Nuclear Information System (INIS)

    Rowden, M.A.; Mason, E.A.; Gilinsky, V.; Kennedy, R.T.

    1976-01-01

    The paper contains the text of a decision of the US NRC that the export of the ASCO nuclear power unit II to Spain would not be inimical to the common defense and security of the United States, so that there are no objections to issue the license to Westinghouse Electric Corporation. Furthermore the paper contains the dissenting opinion of Commissioner Gilinsky. (HP) [de

  3. New Westinghouse correlation WRB-1 for predicting critical heat flux in rod bundles with mixing vane grids

    International Nuclear Information System (INIS)

    Motley, F.E.; Hill, K.W.; Cadek, F.F.; Shefcheck, J.

    1976-07-01

    A new critical heat flux (CHF) correlation, based on local fluid conditions, has been developed from Westinghouse rod bundle data. This correlation applies to both 0.422 inch and 0.374 inch rod O.D. geometries. It accounts for typical cell and thimble cell effects, uniform and non-uniform heat flux profiles, variations in rod heated length and in grid spacing. The correlation predicts CHF for 1147 data points with a sample mean and standard deviation of measured-to-predicted heat flux ratio of 1.0043 and 0.0873, respectively. It was concluded that to meet the reactor design criterion the minimum DNBR should be 1.17

  4. Sensitivity analysis for thermo-hydraulics model of a Westinghouse type PWR. Verification of the simulation results

    Energy Technology Data Exchange (ETDEWEB)

    Farahani, Aref Zarnooshe [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering, Science and Research Branch; Yousefpour, Faramarz [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Hoseyni, Seyed Mohsen [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Basic Sciences; Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Young Researchers and Elite Club

    2017-07-15

    Development of a steady-state model is the first step in nuclear safety analysis. The developed model should be qualitatively analyzed first, then a sensitivity analysis is required on the number of nodes for models of different systems to ensure the reliability of the obtained results. This contribution aims to show through sensitivity analysis, the independence of modeling results to the number of nodes in a qualified MELCOR model for a Westinghouse type pressurized power plant. For this purpose, and to minimize user error, the nuclear analysis software, SNAP, is employed. Different sensitivity cases were developed by modification of the existing model and refinement of the nodes for the simulated systems including steam generators, reactor coolant system and also reactor core and its connecting flow paths. By comparing the obtained results to those of the original model no significant difference is observed which is indicative of the model independence to the finer nodes.

  5. Analysis of volatile headspace gases sampled by cryogenic traps from Westinghouse Hanford Company Tank 242-C-112 March 1992

    International Nuclear Information System (INIS)

    Lucke, R.B.; Clauss, S.A.

    1993-10-01

    Results are given from gas chromatography/mass spectrometry (GC/MS) analyses of the headspace samples obtained by using cryogenic traps from Westinghouse Hanford Company (WHC) Tank 112-C during the month of March, 1992. Samples were analyzed as received with no sample preparation. Analyses included direct GC/MS for volatile/semivolatile components, and direct GC/MS for ammonia. Purge and trap GC/MS analysis was not done. In addition, aliquots were sent to Karl Pool, Pacific Northwest Laboratory, for hydrogen cyanide analysis by ion chromatography, the results are reported here. All concentrations are reported for the methanol extract solutions. To calculate concentrations in the headspace, the cryo-sampling air volume and the methanol rinse volume must be obtained from cryo-sampling personnel at WHC. Triplicate analyses were done on all samples, and average concentrations and standard deviations are reported. One significant result was that no ammonia was detected

  6. The impact of instrumentation and control requirements on the design changes of the Westinghouse ''NSSS'' of Almaraz, Lemoniz and Asco

    International Nuclear Information System (INIS)

    Gerini, P.M.; Naredo, F.P.; Williams, D.W.

    1978-01-01

    For the nuclear power plants Almaraz, Lemoniz and Asco the NSSS set is supplied by Westinghouse. Purchasing contracts were signed in 1971 and projects design took into account the compliance with the regulatory requirements for licensing, issued and standing that time. Since 1971 licensing regulations have been subjected to a deep revision due to the issue of new standards and guides and revision of other affecting altogether the engineering design of nuclear power plants. This situation was reasonably reflected on several consecutive design revisions for the case of the Almaraz, Lemoniz and Asco Nuclear plants. This impact, from the viewpoint of the instrumentation and control context, and referred to the NSSS is analyzed in the report. In particular, attention is paid to the safeguards actuation logic, testing capability and physical separation criteria as contemplated into the regulatory requirements starting from 1971.(J.E.de C)

  7. Preliminary Performance Data on Westinghouse Electronic Power Regulator Operating on J34-WE-32 Turbojet Engine in Altitude Wind Tunnel

    Science.gov (United States)

    Ketchum, James R.; Blivas, Darnold; Pack, George J.

    1950-01-01

    The behavior of the Westinghouse electronic power regulator operating on a J34-WE-32 turbojet engine was investigated in the NACA Lewis altitude wind tunnel at the request of the Bureau of Aeronautics, Department of the Navy. The object of the program was to determine the, steady-state stability and transient characteristics of the engine under control at various altitudes and ram pressure ratios, without afterburning. Recordings of the response of the following parameters to step changes in power lever position throughout the available operating range of the engine were obtained; ram pressure ratio, compressor-discharge pressure, exhaust-nozzle area, engine speed, turbine-outlet temperature, fuel-valve position, jet thrust, air flow, turbine-discharge pressure, fuel flow, throttle position, and boost-pump pressure. Representative preliminary data showing the actual time response of these variables are presented. These data are presented in the form of reproductions of oscillographic traces.

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

    International Nuclear Information System (INIS)

    Smith, Curtis Lee

    2015-01-01

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

  9. Benchmarking of the PHOENIX-P/ANC [Advanced Nodal Code] advanced nuclear design system

    International Nuclear Information System (INIS)

    Nguyen, T.Q.; Liu, Y.S.; Durston, C.; Casadei, A.L.

    1988-01-01

    At Westinghouse, an advanced neutronic methods program was designed to improve the quality of the predictions, enhance flexibility in designing advanced fuel and related products, and improve design lead time. Extensive benchmarking data is presented to demonstrate the accuracy of the Advanced Nodal Code (ANC) and the PHOENIX-P advanced lattice code. Qualification data to demonstrate the accuracy of ANC include comparison of key physics parameters against a fine-mesh diffusion theory code, TORTISE. Benchmarking data to demonstrate the validity of the PHOENIX-P methodologies include comparison of physics predictions against critical experiments, isotopics measurements and measured power distributions from spatial criticals. The accuracy of the PHOENIX-P/ANC Advanced Design System is demonstrated by comparing predictions of hot zero power physics parameters and hot full power core follow against measured data from operating reactors. The excellent performance of this system for a broad range of comparisons establishes the basis for implementation of these tools for core design, licensing and operational follow of PWR [pressurized water reactor] cores at Westinghouse

  10. Advanced core monitoring technology for WWER reactors

    International Nuclear Information System (INIS)

    Nguyen, T.Q.; Casadei, A.L.; Doshi, P.K.

    1993-01-01

    The Westinghouse BEACON online monitoring system has been developed to provide continuous core monitoring and operational support for pressurized water reactor using movable detectors (fission chamber) and core thermocouples. The basic BEACON core monitoring methodology is described. Traditional WWER reactors use rhodium fixed in-core detectors as the means to provide detailed core power distribution for surveillance purposes. An adapted version of the BEACON advanced core monitoring and support system is described which seems to be, due to the different demand/response requirements, the optimal solution (for routine surveillance and anomaly detection) for WWER reactors with existing fixed in-core detectors. (Z.S.) 4 refs

  11. Technology which led to the westinghouse inherently safe liquid metal reactor

    International Nuclear Information System (INIS)

    Schmidt, J.E.; Coffield, R.D.; Doncals, R.A.; Kalinowski, J.E.; Markley, R.A.

    1985-01-01

    The Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor programs resulted in an understanding of liquid metal reactor behavior that is being used to design inherent safety capability into liquid metal reactors. Technological advances give the same beneficial operating characteristics of conventional liquid metal reactors, however, the addition of inherently safe design features precludes the initiation of hypothetical core disruptive accidents. These innovative features permit inherent safety capability to be demonstrated with more than adequate margins. Also, the variety of inherent safety features provides the designers with options in selecting inherent design features for a specific reactor application

  12. IE Information Notice No. 85-93: Westinghouse Type DS circuit breakers, potential failure of electric closing feature because of broken spring release latch lever

    International Nuclear Information System (INIS)

    Jordan, E.L.

    1992-01-01

    On April 14, 1985, the Westinghouse Nuclear Services Integration Division (NSID) issued Technical Bulletin No. NSID-TB-85-17 advising their customers of a potential malfunction in Westinghouse Type DS Class 1E circuit breakers because of broken spring release latch levers. These electrically operated type DS breakers will not close electrically when the spring release latch lever has been broken off. Twenty-five broken levers have been reported and evaluated. This evaluation shows concentrations of incidents traceable to manufacturing in the following periods of time: early 1975, April 1976, and early 1978. This circuit breaker failure, as discussed, adversely affects the safety function (closing on demand) when the circuit breaker is used in the Engineered Safety Features Systems. However, this failure mode will not affect the safety trip function when it is used in the reactor protection system

  13. Assessment of Westinghouse Hanford Company methods for estimating radionuclide release from ground disposal of waste water at the N Reactor sites

    International Nuclear Information System (INIS)

    1988-09-01

    This report summarizes the results of an independent assessment by Golder Associates, Inc. of the methods used by Westinghouse Hanford Company (Westinghouse Hanford) and its predecessors to estimate the annual offsite release of radionuclides from ground disposal of cooling and other process waters from the N Reactor at the Hanford Site. This assessment was performed by evaluating the present and past disposal practices and radionuclide migration data within the context of the hydrology, geology, and physical layout of the N Reactor disposal site. The conclusions and recommendations are based upon the available data and simple analytical calculations. Recommendations are provided for conducting more refined analyses and for continued field data collection in support of estimating annual offsite releases. Recommendations are also provided for simple operational and structural measures that should reduce the quantities of radionuclides leaving the site. 5 refs., 9 figs., 1 tab

  14. WRAP 2A advanced conceptual design report comments

    International Nuclear Information System (INIS)

    Lamberd, D.L.

    1994-01-01

    This report contains the compilation of the 393 comments that were submitted during the review of the Advanced Conceptual Design Report for the Waste Receiving and Processing Facility Module 2A. The report was prepared by Raytheon Engineers and Constructors, Inc. of Englewood, Colorado for the United States Department of Energy. The review was performed by a variety of organizations identified in the report. The comments were addressed first by the Westinghouse cognizant engineers and then by the Raytheon cognizant engineers, and incorporated into the final issue of the Advanced Conceptual Design Report

  15. Health assessment for Westinghouse Elevator, Cumberland Township, Adams County, Pennsylvania, Region 3. CERCLIS No. PAD043882281. Preliminary report

    Energy Technology Data Exchange (ETDEWEB)

    1989-01-19

    The Westinghouse Elevator site in Adams County, Pennsylvania has been in operation since 1968 as an elevator component manufacturing facility. During the process of elevator cab production, parts were passed through a degreasing and paint phase. Sampling of nearby surface water in 1983 revealed contamination with organic solvents. The environmental contamination on-site consists of 1,1,1-trichloroethane and trichloroethylene in surface sludge, soil, surface water, sediment, and groundwater. In addition, 1,1-dichloroethane and 1,1-dichloroethylene were detected in soil. The environmental contamination off-site consists of 1,1,1-trichloroethane, 1,1-dichloroethylene, 1,1-dichloroethane, and trichloroethylene in residential water supply wells. Contaminated groundwater is of primary importance to the site. Private wells have been found to be contaminated and alternate water supplies have been provided. One public supply well has been decommissioned due to contamination. The site is considered to be of potential public health concern because of the risk to human health caused by the possibility of exposure to hazardous substances via contaminated groundwater, surface water, sediment, and possibly on-site soil.

  16. Westinghouse Hanford Company ALARA year-end report, Calendar Year 1994: Revision 3A, Radiological engineering and ALARA

    International Nuclear Information System (INIS)

    Berglund, O.D.

    1995-06-01

    It has long been the US Department of Energy's (DOE's) Policy that radiation doses should be maintained as far below the dose limits as is reasonably achievable. This policy, known as the ''ALARA Principle of radiation protection,'' maintains that radiation exposures should be maintained as low as reasonably achievable, taking into account social, technical, economic, practical, and public policy considerations. The ALARA Principle is based on the hypothesis that even very low radiation doses carry some risk. As a result, it is not enough to maintain doses at/or slightly below limits; the lower the doses, the lower the risks. Because it is not possible to reduce all doses at DOE facilities to zero, economic and social factors must be considered to determine the optimal level of radiation doses. According to the ALARA Principle, if doses are too high, resources should be well spent to reduce them. At some point, the resources being spent to maintain low doses are exactly balanced by the risks avoided. Reducing doses below this point results in a misallocation of resources; the resources could be spent elsewhere and have a greater positive impact on health and safety. The objective of the Westinghouse Hanford Company (WHC) ALARA/Contamination Control Improvement Project (CCIP) Program is to manage and control exposures (both individual and collective) to the work force, the general public, and the environment to levels as low as is reasonable using the aforementioned ALARA Principle

  17. Westinghouse-Gothic comparisons with passive containment cooling tests using a one-to-ten-scale test facility

    International Nuclear Information System (INIS)

    Kennedy, M.D.; Woodcock, J.; Wright, R.F.; Gresham, J.A.

    1996-01-01

    The Heavy Water Reactor Facility is equipped with a passive cooling system to provide long-term decay heat removal during postulated beyond-design-basis accidents. The passive containment cooling system (PCCS) consists of an annular space between the steel containment vessel and the concrete shield building and optimized inlet and chimney designs. The design, analysis, and regulatory acceptance of a plant with PCCS requires an understanding of the external convective and radiative heat transfer phenomena, as well as the internal distributions of noncondensable gases. The internal distribution of noncondensable gases has a strong effect on the resistance to condensation heat transfer and therefore affects the wall temperature distribution applied to the external channel. To evaluate these phenomena, a test facility having a scale of approximately one to ten, known as the large-scale test, was constructed, and several series of tests were performed. Test results have been used to validate the Westinghouse-GOTHIC (WGOTHIC) computer code. A comparison of WGOTHIC predictions and test results has been completed. This paper shows that mixed-convection models applied to the interior and exterior surfaces as well as a heat and mass transfer analogy for internal condensation provides good comparison to test results. An axial distribution of noncondensables within the test vessel is also predicted

  18. Test and evaluation report for Westinghouse Hanford Company's Hedgehog Shielded Container, Docket 94-39-7A, Type A container

    International Nuclear Information System (INIS)

    Kelly, D.L.

    1995-01-01

    This report documents the US Department of Transportation Specification 7A Type A (DOT-7A) compliance test results of the Westinghouse Hanford Company Hedgehog Shielded Container. The Hedgehog packaging configurations provide primary and secondary containment. The packaging configurations tested consisted of an internal bottle, varying in size. Testing showed that the bottles are not required for the packaging to pass Type A requirements, with the exception of the 1-liter version, in which the polyvinyl chloride (PVC)-coated glass bottle used in testing is considered a part of the containment system. The packaging configurations were evaluated and tested in February 1995. The packaging configurations described in this report are designed to ship Type A quantities of radioactive materials, normal form. Contents may be in solid or liquid form. Liquids may have a specific gravity ≤2. The solid versions would allow the shipment of normal or special form solids. The solid materials would be limited in weight--to include packaging--to the gross weight of the as-tested liquids and bottles. The packaging configurations described in this document may be transported by air, and they meet the applicable International Air Transport Association/International Civil Aviation Organization (IATA/ICAO) Dangerous Goods Regulations in addition to the DOT-7A requirements

  19. A New Neutron Multiplicity Counter for the Measurement of Impure Plutonium Metal at Westinghouse Savannah River Site

    International Nuclear Information System (INIS)

    Baker, L.B.; Faison, D.M.; Langner, D.G.; Sweet, M.R.; Salazar, S.D.; Kroncke, K.E.

    1998-07-01

    A new neutron multiplicity counter has been designed, fabricated, characterized, and installed for use in the assay of impure plutonium metal buttons from the FB-Line at the Westinghouse Savannah River Site (WSRS). This instrument incorporates the performance characteristics of the Pyrochemical or In-plant Multiplicity Counter with the package size of the Plutonium Scrap Multiplicity Counter. In addition, state-of-the art features such as the de-randomizer circuit and separate ring outputs have been added. The counter consists of 113, 71 cm active length 3He tubes in a polyethylene moderator. Its efficiency for 252Cf is 57.8 percent, the highest of any multiplicity counter to date. Its die-away time is 50.4 ms and its deadtime is 50 ns. In this paper we will present the characterization data for the counter and the results of preliminary metal measurements at WSRS. We will also discuss the new challenges the impure metal buttons from FB-Line are presenting to the multiplicity counting technique

  20. OPTIMIZATION OF ADVANCED FILTER SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    R.A. Newby; G.J. Bruck; M.A. Alvin; T.E. Lippert

    1998-04-30

    Reliable, maintainable and cost effective hot gas particulate filter technology is critical to the successful commercialization of advanced, coal-fired power generation technologies, such as IGCC and PFBC. In pilot plant testing, the operating reliability of hot gas particulate filters have been periodically compromised by process issues, such as process upsets and difficult ash cake behavior (ash bridging and sintering), and by design issues, such as cantilevered filter elements damaged by ash bridging, or excessively close packing of filtering surfaces resulting in unacceptable pressure drop or filtering surface plugging. This test experience has focused the issues and has helped to define advanced hot gas filter design concepts that offer higher reliability. Westinghouse has identified two advanced ceramic barrier filter concepts that are configured to minimize the possibility of ash bridge formation and to be robust against ash bridges should they occur. The ''inverted candle filter system'' uses arrays of thin-walled, ceramic candle-type filter elements with inside-surface filtering, and contains the filter elements in metal enclosures for complete separation from ash bridges. The ''sheet filter system'' uses ceramic, flat plate filter elements supported from vertical pipe-header arrays that provide geometry that avoids the buildup of ash bridges and allows free fall of the back-pulse released filter cake. The Optimization of Advanced Filter Systems program is being conducted to evaluate these two advanced designs and to ultimately demonstrate one of the concepts in pilot scale. In the Base Contract program, the subject of this report, Westinghouse has developed conceptual designs of the two advanced ceramic barrier filter systems to assess their performance, availability and cost potential, and to identify technical issues that may hinder the commercialization of the technologies. A plan for the Option I, bench

  1. The importance of collaboration in the advancement of current and next generation reactors

    International Nuclear Information System (INIS)

    Jackson, Kate; Goossen, John; Anness, Mike; Meston, Tom

    2010-01-01

    The sections of the contribution are as follows: Tradition of innovation. Growing demand for nuclear power; Collaboration drivers; Responses. Knowledge transfer and management is critical. What kind of focus? Equipment reliability. Advanced repair, replacement and construction approaches. Materials. Plant safety margins. Spent fuel management. Examples of European collaboration. Zorita materials examination. Collaboration in the development of next generation reactors; Westinghouse R and D priorities; A look to the future. (P.A.)

  2. Status of developing advanced PWR in Japan

    International Nuclear Information System (INIS)

    Iida, Yotaro

    1982-01-01

    During past eleven years since the first PWR power plant, Mihama Unit 1 of Kansai Electric Power Co., started the commercial operation in 1970, Mitsubishi Heavy Industries has endeavored to improve PWR technologies on the basis of the advice from electric power companies and the technical information to overcome difficulties in PWR power plants. Now, the main objective is to improve the overall plant performance, and the rate of operation of Japanese PWR power plants has significantly risen. The improvement of the reliability, the shortening of regular inspection period and the reduction of radioactive waste handling were attempted. In view of the satisfactory operational experience of Westinghouse type PWRs, the basic reactor concept has not been changed so far. Mitsubishi and Westinghouse reached basic agreement in August, 1981, to develop a spectral shift type large capacity reactor as the advanced PWRs for Japan. This type of PWRs hab higher degree of freedom for extended fuel cycle operation and enhances the advantage of entire fuel cycle economy, particularly the significant reduction of uranium use. The improved neutron economy is attainable by reducing neutron loss, and the core design with low power density and the economical use of plutonium are advantageous for the fuel cycle economy. (Kako, I.)

  3. Radioactive waste isolation in salt: peer review of Westinghouse Electric Corporation's report on reference conceptual designs for a repository waste package

    International Nuclear Information System (INIS)

    Rote, D.M.; Hull, A.B.; Was, G.S.; Macdonald, D.D.; Wilde, B.E.; Russell, J.E.; Kruger, J.; Harrison, W.; Hambley, D.F.

    1985-10-01

    This report documents the findings of the peer panel constituted by Argonne National Laboratory to review Region A of Westinghouse Electric Corporation's report entitled Waste Package Reference Conceptual Designs for a Repository in Salt. The panel determined that the reviewed report does not provide reasonable assurance that US Nuclear Regulatory Commission (NRC) requirements for waste packages will be met by the proposed design. It also found that it is premature to call the design a ''reference design,'' or even a ''reference conceptual design.'' This review report provides guidance for the preparation of a more acceptable design document

  4. Analysis of a hot-leg small break loss-of-coolant accident in a three-loop westinghouse pressurized water reactor plant

    International Nuclear Information System (INIS)

    Peterson, C.E.; Chexal, V.K.; Clements, T.B.

    1985-01-01

    The RETRAN-02 computer code was used to perform a best-estimate analysis of a 7.52-cm-diam hotleg break in a three-loop Westinghouse pressurized water reactor. This break size produced a net primary coolant mass depletion through the early portion of the transient. The primary system started to refill only after the accumulator valves opened. As the primary system refilled, there were extreme temperature differentials around the system with cold, denser fluid collecting at the lower elevations and two-phase fluid at higher elevations

  5. RELAP/MOD1.5 analysis of steam line break transients for a 3-loop and a 4-loop Westinghouse nuclear steam supply system

    International Nuclear Information System (INIS)

    Peeler, G.B.; McDonald, T.A.; Kennedy, M.F.

    1984-01-01

    RELAP/MOD1.5 (Cycle 31 and 34) calculations were made to assess the assumptions used by Westinghouse (W) to analyze mainsteam line break transients. Models of a W 3-loop and 4-loop nuclear steam supply system were used. Sensitivity studies were performed to determine the effect of the availability of offsite power, break size and initial core power. Comparison with W results indicated that if the assumptions used by W are replicated within the RELAP5 framework, then the W methodology for prediction of the Nuclear Steam Supply System (NSSS) response is conservative for steam line break transients

  6. Radioactive waste isolation in salt: peer review of Westinghouse Electric Corporation's report on reference conceptual designs for a repository waste package

    Energy Technology Data Exchange (ETDEWEB)

    Rote, D.M.; Hull, A.B.; Was, G.S.; Macdonald, D.D.; Wilde, B.E.; Russell, J.E.; Kruger, J.; Harrison, W.; Hambley, D.F.

    1985-10-01

    This report documents the findings of the peer panel constituted by Argonne National Laboratory to review Region A of Westinghouse Electric Corporation's report entitled Waste Package Reference Conceptual Designs for a Repository in Salt. The panel determined that the reviewed report does not provide reasonable assurance that US Nuclear Regulatory Commission (NRC) requirements for waste packages will be met by the proposed design. It also found that it is premature to call the design a ''reference design,'' or even a ''reference conceptual design.'' This review report provides guidance for the preparation of a more acceptable design document.

  7. Advanced passive technology: A global standard for nuclear plant requirements

    International Nuclear Information System (INIS)

    Novak, V.

    1994-01-01

    Since 1984, Westinghouse has been developing AP8OO, a 800 MW, two-loop advanced passive plant, in response to an initiative established by the Electric Power Research Institute (EPRI) and the U.S. Department of Energy' (DOE). The preliminary design was cornpleved in 1989. AP6OO's Standard Safety Analysis and Probabilistic Risk analysis Reports were submitted to the U.S. Nuclear Regulatory Commission for design certification in 1992. Design simplification is the key strategy behind the AP6OO. The basic technical concept Of simplification has resulted in a simplified reactor coolant systems, simplified plant systems, a simplified plant arrangement, reduced number of components, simplified operation and maintenance

  8. Gas-cooled reactors for advanced terrestrial applications

    International Nuclear Information System (INIS)

    Kesavan, K.; Lance, J.R.; Jones, A.R.; Spurrier, F.R.; Peoples, J.A.; Porter, C.A.; Bresnahan, J.D.

    1986-01-01

    Conceptual design of a power plant on an inert gas cooled nuclear coupled to an open, air Brayton power conversion cycle is presented. The power system, called the Westinghouse GCR/ATA (Gas-Cooled Reactors for Advanced Terrestrial Applications), is designed to meet modern military needs, and offers the advantages of secure, reliable and safe electrical power. The GCR/ATA concept is adaptable over a range of 1 to 10 MWe power output. Design descriptions of a compact, air-transportable forward base unit for 1 to 3 MWe output and a fixed-base, permanent installation for 3 to 10 MWe output are presented

  9. Human Performance Westinghouse Program

    International Nuclear Information System (INIS)

    Garcia Gutierrez, A.; Gil, C.

    2010-01-01

    The objective of the Program consists in the excellence actuation, achieving the client success with a perfect realisation project. This program consists of different basic elements to reduce the human mistakes: the HuP tools, coaching, learning clocks and iKnow website. There is, too, a document file to consult and practice. All these elements are expounded in this paper.

  10. Regulatory Considerations for the Long Term Cooling Safe Shutdown Requirements of the Passive Residual Heat Removal Systems in Advanced Reactors

    International Nuclear Information System (INIS)

    Sim, S. K.; Bae, S. H.; Kim, Y. S.; Hwang, Min Jeong; Bang, Young Seok; Hwang, Taesuk

    2016-01-01

    USNRC approved safe shutdown at 215.6 .deg. C for a safe and long term cooling state for the redundant passive RHRSs by SECY-94-084. USNRC issued COLA(Combined Construction and Operating License) for the Levy County NP Unit-1/2 for the AP1000 passive RHRSs in 2014. Korea Hydro and Nuclear Power(KHNP) is developing APR+ and adopted Passive Auxiliary Feedwater System(PAFS) as a new passive RHRS design. Korea Institute of Nuclear Safety(KINS) has been developing regulatory guides for the advanced safety design features of the advanced ALWRs which has plan to construct in near future in Korea[5]. Safety and regulatory issues as well as the safe shut down requirements of the passive RHRS are discussed and considerations in developing regulatory guides for the passive RHRS are presented herein. Passive RHRSs have been introduced as new safety design features for the advanced reactors under development in Korea. These passive RHRSs have potential advantages over existing active RHRS, however, their functions are limited due to inherent ability of passive heat removal processes. It is high time to evaluate the performance of the passive PRHRs and develop regulatory guides for the safety as well as the performance analyses of the passive RHRS

  11. Establishment of the operating procedure to prevent boron precipitation during Post-LOCA long term cooling for Korean Westinghouse 3-loop NPPs

    International Nuclear Information System (INIS)

    Choi, Han Rim; Kwon, Tae Soon; Ban, Chang Hwan; Jeong, Jae Hoon; Lee, Young Jin.

    1996-11-01

    During post-LOCA LTC the increase of the excess reactivity for the extended fuel cycle should require increasing the RWST boron concentration in order to ensure core subcritical state. To quantify the concentration increment, the calculation methods was developed for the post-LOCA RCS/Sump mixed mean boron concentration, which applied for Kori 3 and 4 and Ulchin 1 and 2 of the Westinghouse 3-loop nuclear power plants in Korean. From the calculation results, the minimum boric acid concentrations increased of the RWST and accumulator were determined consideration of the convenient operation for operator on reloading. Boric acid concentrations of the RWST and the accumulators for Westinghouse 3-loop type plants were increased to meet the post-LOCA shutdown requirement for the long life cycles from 12 months to 18 months. To maintain LTC capability following a LOCA, the switchover time is examined using boron code of prevent the boron precipitation in the reactor core with the increased boron concentrations. The analysis results showed that hot leg recirculation switchover times were shortened to 7.5 hours from 24 hours after the initiation of LOCA for Kori 3 and 4 and 8 hours from 18 hours for Ulchin 1 and 2, respectively. The flow path in the mode J for Kori 3 and 4 was recommended to realign to the simultaneous recirculation of both hot and cold legs from the cold leg recirculation, as done by Ulchin 1 and 2. (author). 2 tabs., 12 figs., 13 refs

  12. IE Information Notice No. 85-18, Supplement 1: Failures of undervoltage output circuit boards in the Westinghouse-designed solid state protection system

    International Nuclear Information System (INIS)

    Rossi, C.E.

    1992-01-01

    The US Nuclear Regulatory Commission (NRC) is issuing this information notice supplement to alert addressees to continuing problems associated with the undervoltage (UV) output circuit boards (driver cards) in the solid state protection system (SSPS) designed by the Westinghouse Electric Corporation (Westinghouse). On June 3, 1991, the Shearon Harris Nuclear Power Plant, Unit 1, (Harris) experienced an automatic reactor trip from 100 percent power on a spurious low reactor coolant system loop flow signal. The signal was generated as a result of a surveillance test being performed on one of three loop flow transmitters. The licensee attributed the spurious signal to both procedural inadequacies and personnel error. A control room operator verified that all control rods had fully inserted following the trip signal and that reactor power was properly decreasing. However, about 22 seconds after the automatic trip signal was generated, operators discovered that the ''A'' reactor trip breaker (RTB) had not opened. The RTB was manually opened using the reactor trip switch on the main control board. Subsequent analyses are discussed

  13. TRAC-PF1/MOD2 best-estimate analysis of a large-break LOCA in a 15 x 15 generic four-loop Westinghouse nuclear power plant

    International Nuclear Information System (INIS)

    Spore, J.W.; Lin, J.C.; Schnurr, N.M.; White, J.R.; Cappiello, M.C.

    1992-01-01

    Calculations of a large-break loss-of-coolant accident (LOCA) in a 15 x 15 generic four-loop Westinghouse nuclear power plant with both the TRAC-PF1/MOD1 and TRAC-PF1/MOD2 computer codes will be presented. The Transient Reactor Analysis Code (TRAC) has been developed by Los Alamos National Laboratory to provide advanced best-estimate simulations of real postulated transients in pressurized light-water reactors (LWRs) and for many related thermal-hydraulic facilities. The latest released version of TRAC is TRAC-PF1/MOD2. Significant improvements and enhancements over the MOD1 version were implemented in the MOD2 heat-transfer and constitutive models. One of the most significant improvements in the MOD2 code has been the implementation of the two-step numerics method in the three-dimensional components, which can significantly reduce run times for long, slow transients. A very important area of improvement has been in the reflood heat-transfer models. Developmental assessment results (i.e., code comparisons with experimental data) will be discussed for several separate-effects and integral test, including analysis of the Upper Plenum Test Facility (UPTF), the Cylindrical Core Test Facility (CCTF), and the Loss-of-Fluid Test Facility (LOFT). The assessment results provide information on the anticipated accuracy for the best-estimate models in the MOD2 computer code. The MOD1 to MOD2 comparison will provide an estimate for the effect of improved heat-transfer models on predicted peak cladding temperatures

  14. Evaluation of the PRHRS Performance Degradation due to Non-Condensable Gas for the Small and Medium Reactor using MARS-KS code

    International Nuclear Information System (INIS)

    Kim, Sook Kwan; Sim, Suk Ku; Park, Ju Yeop; Seol, Kwang Won; Ryu, Yong Ho

    2011-01-01

    The effect of non-condensable gas on the performance of PRHRS (Passive Residual Heat Removal System) of the Small and Medium Reactor(SMR) was evaluated during a loss of flow event. Since the TMI accident in 1979, the passive systems have been considered in the advanced reactors as a feature of design improvement because the passive system simplifies the system and thus increases the reliability of the system. The Westinghouse received the design certification from the USNRC for the AP600 and AP1000 passive type pressurized water reactors. The APR+ under development by KEPCO considers the use of PAFS (Passive Auxiliary Feedwater System). And the PRHRS is adopted as a passive secondary heat removal system for the SMART (System-integrated Modular Advanced ReacTor)

  15. Implementation, capabilities, and benchmarking of Shift, a massively parallel Monte Carlo radiation transport code

    International Nuclear Information System (INIS)

    Pandya, Tara M.; Johnson, Seth R.; Evans, Thomas M.; Davidson, Gregory G.; Hamilton, Steven P.; Godfrey, Andrew T.

    2015-01-01

    This paper discusses the implementation, capabilities, and validation of Shift, a massively parallel Monte Carlo radiation transport package developed and maintained at Oak Ridge National Laboratory. It has been developed to scale well from laptop to small computing clusters to advanced supercomputers. Special features of Shift include hybrid capabilities for variance reduction such as CADIS and FW-CADIS, and advanced parallel decomposition and tally methods optimized for scalability on supercomputing architectures. Shift has been validated and verified against various reactor physics benchmarks and compares well to other state-of-the-art Monte Carlo radiation transport codes such as MCNP5, CE KENO-VI, and OpenMC. Some specific benchmarks used for verification and validation include the CASL VERA criticality test suite and several Westinghouse AP1000 ® problems. These benchmark and scaling studies show promising results

  16. Study on natural circulation flow under reactor cavity flooding condition in advanced PWRs

    International Nuclear Information System (INIS)

    Tao Jun; Yang Jiang; Cao Jianhua; Lu Xianghui; Guo Dingqing

    2015-01-01

    Cavity flooding is an important severe accident management measure for the in-vessel retention of a degraded core by external reactor vessel cooling in advanced PWRs. A code simulation study on the natural circulation flow in the gap between the reactor vessel wall and insulation material under cavity flooding condition is performed by using a detailed mechanistic thermal-hydraulic code package RELAP 5. By simulating of an experiment carried out for studying the natural circulation flow for APR1400 shows that the code is applicable for analyzing the circulation flow under this condition. The analysis results show that heat removal capacity of the natural circulation flow in AP1000 is sufficient to prevent thermal failure of the reactor vessel under bounding heat load. Several conclusions can be drawn from the sensitivity analysis. Larger coolant inlet area induced larger natural circulation flow rate. The outlet should be large enough and should not be submerged by the cavity water to vent the steam-water mixture. In the implementation of cavity flooding, the flooding water level should be high enough to provide sufficient natural circulation driven force. (authors)

  17. Advanced light water reactors for the nineties

    International Nuclear Information System (INIS)

    Ross, F.A.; Sugnet, W.R.

    1987-01-01

    The EPRI/Industry advanced light water reactor (ALWR) program and the US Department of Energy ALWR program are closely coordinated to meet the common objective which is the availability of improved and simplified light water reactor plants that may be ordered in the next decade to meet new or replacement capacity requirements. The EPRI/Industry objectives, program participants, and foreign participants, utility requirements document, its organization and content, small plant conceptual design program, the DOE ALWR program, design verification program, General Electric ABWR design features, Combustion Engineering system design, mid-size plant development, General Electric SBWR objectives, Westinghouse/Burns and Roe design objectives, construction improvement, and improved instrumentation and control are discussed in the paper

  18. High-inertia hermetically sealed main coolant pump for next generation passive nuclear power plants

    International Nuclear Information System (INIS)

    Kujawski, Joseph M.; Nair, Bala R.; Vijuk, Ronald P.

    2003-01-01

    The main coolant pump for the Westinghouse AP1000 advanced passive nuclear power plant represents a significant scale-up in power, flow capacity, and physical size from its predecessor designed for the smaller AP600 power plant. More importantly, the AP1000 pump incorporates several innovative features that contribute to improved efficiency, operational reliability, and plant safety. The features include an internals design which provides the highest hydraulic efficiency achieved in commercial nuclear power plant applications. Another feature is the use of a distributed inertial mass system in the rotating assembly to develop the high rotational inertia to meet the extended system flow coastdown requirement for core heat removal in the event of loss of power to the pumps. This advanced canned motor pump also incorporates the latest development in higher operating voltage, providing plant designers with the ability to eliminate plant transformers and operate directly on the site electrical bus in many cases. The salient features of the pump design and performance data are presented in this paper. (author)

  19. OPTIMIZATION OF ADVANCED FILTER SYSTEMS; TOPICAL

    International Nuclear Information System (INIS)

    R.A. Newby; G.J. Bruck; M.A. Alvin; T.E. Lippert

    1998-01-01

    Reliable, maintainable and cost effective hot gas particulate filter technology is critical to the successful commercialization of advanced, coal-fired power generation technologies, such as IGCC and PFBC. In pilot plant testing, the operating reliability of hot gas particulate filters have been periodically compromised by process issues, such as process upsets and difficult ash cake behavior (ash bridging and sintering), and by design issues, such as cantilevered filter elements damaged by ash bridging, or excessively close packing of filtering surfaces resulting in unacceptable pressure drop or filtering surface plugging. This test experience has focused the issues and has helped to define advanced hot gas filter design concepts that offer higher reliability. Westinghouse has identified two advanced ceramic barrier filter concepts that are configured to minimize the possibility of ash bridge formation and to be robust against ash bridges should they occur. The ''inverted candle filter system'' uses arrays of thin-walled, ceramic candle-type filter elements with inside-surface filtering, and contains the filter elements in metal enclosures for complete separation from ash bridges. The ''sheet filter system'' uses ceramic, flat plate filter elements supported from vertical pipe-header arrays that provide geometry that avoids the buildup of ash bridges and allows free fall of the back-pulse released filter cake. The Optimization of Advanced Filter Systems program is being conducted to evaluate these two advanced designs and to ultimately demonstrate one of the concepts in pilot scale. In the Base Contract program, the subject of this report, Westinghouse has developed conceptual designs of the two advanced ceramic barrier filter systems to assess their performance, availability and cost potential, and to identify technical issues that may hinder the commercialization of the technologies. A plan for the Option I, bench-scale test program has also been developed based

  20. Exxon Nuclear Company ECCS evaluation of a 2-loop Westinghouse PWR with dry containment using the ENC WREM-II ECCS model. Large break example problem

    International Nuclear Information System (INIS)

    Krajicek, J.E.

    1977-01-01

    This document is presented as a demonstration of the ENC WREM-II ECCS model calculational procedure applied to a Westinghouse 2-loop PWR with a dry containment (R. E. Ginna plant, for example). The hypothesized Loss-of-Coolant Accident (LOCA) investigated was a split break with an area equal to twice the pipe cross-sectional area. The break was assumed to occur in one pump discharge pipe (DECLS break). The analyses involved calculations using the ENC WREM-II model. The following codes were used: RELAP4-EM/ENC26A for blowdown and hot channel analyses, RELAP4-EM FLOOD/ENC26A for core reflood analysis, CONTEMPT LT/22 modified for containment backpressure analysis, and TOODEE2/APR77 for heatup analysis

  1. TASS code topical report. V.2 TASS code validation report for the non-LOCA transient analysis of the CE and Westinghouse type plants

    International Nuclear Information System (INIS)

    Sim, Suk K.; Chang, W. P.; Kim, K. D.; Lee, S. J.; Kim, H. C.; Yoon, H. Y.

    1997-02-01

    The development of TASS 1.0 code has been completed and validated its capability in applying for the licensing transient analyses of the CE and Westinghouse type operating reactors as well as the PWR plants under construction in Korea. The validation of the TASS 1.0 code has been achieved through the comparison calculations of the FSAR transients, loss of AC power transient plant data, load rejection and startup test data for the reference plants as well as the BETHSY loop steam generator tube rupture test data. TASS 1.0 calculation agrees well with the best FSAR transient and shows its capability in simulating plant transient analyses. (author). 12 refs., 32 tabs., 132 figs

  2. Estimated airborne release of plutonium from Westinghouse Cheswick site as a result of postulated damage from severe wind and seismic hazard

    International Nuclear Information System (INIS)

    Mishima, J.; Schwendiman, L.C.; Ayer, J.E.

    1979-06-01

    The potential airborne releases of plutonium (source terms) from postulated damage sustained by the Westinghouse Plutonium Fuel Development Laboratories at the Cheswick site in Pennsylvania as a result of various levels of wind and seismic hazard are estimated. The source terms are based on damage scenarios originated by other specialists and range up to 260 mph for wind hazard and in excess of 0.39 g ground acceleration for seismic hazard. The approaches and factors used to estimate the source terms (inventories of dispersible materials at risk, damage levels and ratios, fractional airborne releases of dispersible materials under stress, atmosphere exchange rates, and source term ranges) are discussed. Source term estimates range from less than 10 -7 g plutonium to greater than 130 g plutonium over a four-day period

  3. Operation and maintenance manual addendum to Westinghouse OMM-051-00-005 for intermediate-size inducer pump (ISIP) Model 266

    International Nuclear Information System (INIS)

    1979-01-01

    This Addendum A to the Westinghouse Operation and Maintenance Manual OMM-051-00-005 contains additions, changes and deletions that modify data for the Liquid Metal Coolant Pump, Model LMP-1 to data applicable to the Rockwell International Intermediate-Size Inducer Pump (ISIP), Model 266. The major modifications on the ISIP are the new impeller/inducer assembly, diffuser vanes and new securing hardware. The paragraphs affected by Addendum A use the numbering system used in OMM-051-00-005 except each paragraph number is prefixed with an A preceding the paragraph number signifying that the paragraph is changed to conform with the ISIP configuration. Paragraphs within OMM-051-00-005 not identified in Addendum A remain unchanged with the data still valid and useable

  4. Comparison of the APA-H (Westinghouse) calculations with the operational data for ZpNPP unit 3 cycles 16-19

    International Nuclear Information System (INIS)

    Abdullayev, A. M.; Gorbachenko, O. V.; Ignatchenko, A.I.; Maryokhin, S.V.; Zhukov, A. I.

    2007-01-01

    The computer simulation of ZpNPP Unit 3 (WWER-1000) Cycles 16-19 core depletion has been performed on the basis of the operational data. The changes in reactor heat rate, lead bank position and inlet temperature during the core operation have been taken into account. These calculations were performed by using Westinghouse APA-H (ALPHA/PHOENIX/ ANC-H) code system. The main objectives of the calculations were the comparison with operational data for core loading with TVS-M (Cycle 16) and transition core loading with TVS-A (Cycle 17-19). The calculation results were compared with the results of Critical boric acid concentration vs. Cycle Burnup measurements and Start up Physics Test measurements (at HZP, BOC, NoXe core conditions). Additionally, the comparison between the results of assemblies power calculation performed by ANC-H and BIPR-7A codes is presented (Authors)

  5. Experimental validation of CASMO-4E and CASMO-5M for radial fission rate distributions in a westinghouse SVEA-96 Optima2 BWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Grimm, P.; Perret, G. [Paul Scherrer Inst., CH-5232 Villigen PSI (Switzerland)

    2012-07-01

    Measured and calculated radial total fission rate distributions are compared for the three axial sections of a Westinghouse SVEA-96 Optima2 BWR fuel assembly, comprising 96, 92 and 84 fuel rods, respectively. The measurements were performed on a full-size fuel assembly in the PROTEUS zero-power experimental facility. The measured fission rates are compared to the results of the CASMO-4E and CASMO-5M fuel assembly codes. Detailed measured geometrical data were used in the models, and effects of the surrounding zones of the reactor were taken into account by correction factors derived from MCNPX calculations. The results of the calculations agree well with those of the experiments, with root-mean-square deviations between 1.2% and 1.5% and maximum deviations of 3-4%. The quality of the predictions by CASMO-4E and CASMO-5M is comparable. (authors)

  6. Projects of Modifications of design for mitigation of accidents outside the design Bases on nuclear Central PWR Siemens-KWU and Westinghouse; Proyectos de Modificaciones de Sieno para Mitigacion de Accidentes fuera de la Bases de Diseno en Centrales Nucleares PWR Siemens-KWU y Westinghouse

    Energy Technology Data Exchange (ETDEWEB)

    Dominguez Gonzalez, G.; Cano Rodriguez, L. A.; Arguello Tara, A.

    2014-07-01

    Following the accident at the Japanese Fukushima-Daiichi NPP, the different regulators of nuclear power generation have required numerous reports regarding the evaluation and modification of the capacity of the plants to face accidents with severities beyond that established in their Design Bases. Under this new scenario, with multiple new demands and commitments, EA has carried out the required works for the implementation of strategies to mitigate the consequences of beyond Design Basis accidents for utilities owning Siemens-KWU and Westinghouse PWR nuclear power plants. (Author)

  7. ADVANCED TURBINE SYSTEMS PROGRAM

    Energy Technology Data Exchange (ETDEWEB)

    Gregory Gaul

    2004-04-21

    Natural gas combustion turbines are rapidly becoming the primary technology of choice for generating electricity. At least half of the new generating capacity added in the US over the next twenty years will be combustion turbine systems. The Department of Energy has cosponsored with Siemens Westinghouse, a program to maintain the technology lead in gas turbine systems. The very ambitious eight year program was designed to demonstrate a highly efficient and commercially acceptable power plant, with the ability to fire a wide range of fuels. The main goal of the Advanced Turbine Systems (ATS) Program was to develop ultra-high efficiency, environmentally superior and cost effective competitive gas turbine systems for base load application in utility, independent power producer and industrial markets. Performance targets were focused on natural gas as a fuel and included: System efficiency that exceeds 60% (lower heating value basis); Less than 10 ppmv NO{sub x} emissions without the use of post combustion controls; Busbar electricity that are less than 10% of state of the art systems; Reliability-Availability-Maintainability (RAM) equivalent to current systems; Water consumption minimized to levels consistent with cost and efficiency goals; and Commercial systems by the year 2000. In a parallel effort, the program was to focus on adapting the ATS engine to coal-derived or biomass fuels. In Phase 1 of the ATS Program, preliminary investigators on different gas turbine cycles demonstrated that net plant LHV based efficiency greater than 60% was achievable. In Phase 2 the more promising cycles were evaluated in greater detail and the closed-loop steam-cooled combined cycle was selected for development because it offered the best solution with least risk for achieving the ATS Program goals for plant efficiency, emissions, cost of electricity and RAM. Phase 2 also involved conceptual ATS engine and plant design and technology developments in aerodynamics, sealing

  8. Development of 3D pseudo pin-by-pin calculation methodology in ANC

    International Nuclear Information System (INIS)

    Zhang, B.; Mayhue, L.; Huria, H.; Ivanov, B.

    2012-01-01

    Advanced cores and fuel assembly designs have been developed to improve operational flexibility, economic performance and further enhance safety features of nuclear power plants. The simulation of these new designs, along with strong heterogeneous fuel loading, have brought new challenges to the reactor physics methodologies currently employed in the industrial codes for core analyses. Control rod insertion during normal operation is one operational feature in the AP1000 R plant of Westinghouse next generation Pressurized Water Reactor (PWR) design. This design improves its operational flexibility and efficiency but significantly challenges the conventional reactor physics methods, especially in pin power calculations. The mixture loading of fuel assemblies with significant neutron spectrums causes a strong interaction between different fuel assembly types that is not fully captured with the current core design codes. To overcome the weaknesses of the conventional methods, Westinghouse has developed a state-of-the-art 3D Pin-by-Pin Calculation Methodology (P3C) and successfully implemented in the Westinghouse core design code ANC. The new methodology has been qualified and licensed for pin power prediction. The 3D P3C methodology along with its application and validation will be discussed in the paper. (authors)

  9. Plant maintenance and advanced reactors issue, 2008

    Energy Technology Data Exchange (ETDEWEB)

    Agnihotri, Newal [ed.

    2009-09-15

    The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Technologies of national importance, by Tsutomu Ohkubo, Japan Atomic Energy Agency, Japan; Modeling and simulation advances brighten future nuclear power, by Hussein Khalil, Argonne National Laboratory, Energy and desalination projects, by Ratan Kumar Sinha, Bhabha Atomic Research Centre, India; A plant with simplified design, by John Higgins, GE Hitachi Nuclear Energy; A forward thinking design, by Ray Ganthner, AREVA; A passively safe design, by Ed Cummins, Westinghouse Electric Company; A market-ready design, by Ken Petrunik, Atomic Energy of Canada Limited, Canada; Generation IV Advanced Nuclear Energy Systems, by Jacques Bouchard, French Commissariat a l'Energie Atomique, France, and Ralph Bennett, Idaho National Laboratory; Innovative reactor designs, a report by IAEA, Vienna, Austria; Guidance for new vendors, by John Nakoski, U.S. Nuclear Regulatory Commission; Road map for future energy, by John Cleveland, International Atomic Energy Agency, Vienna, Austria; and, Vermont's largest source of electricity, by Tyler Lamberts, Entergy Nuclear Operations, Inc. The Industry Innovation article is titled Intelligent monitoring technology, by Chris Demars, Exelon Nuclear.

  10. NRC Information No. 91-48: False certificates of conformance provided by Westinghouse Electric Supply Company for refurbished commercial-grade circuit breakers

    International Nuclear Information System (INIS)

    Rossi, C.E.

    1992-01-01

    In April 1988, WESCO of Albany, New York, supplied 250 MCCBs (molded-case circuit breakers) to Spectrum of Schenectady, New York. Spectrum dedicated these commercial-grade items on the basis of independent testing and the certificates of conformance (CoCs) it received from WESCO. During receipt inspection testing, Peach Bottom determined that the MCCBs were not new and had been refurbished. The NRC conducted an inspection and investigation of Spectrum and WESCO in 1988 and 1989. During these efforts, the NRC identified that the MCCBs provided to Peach Bottom were reconditioned and not new equipment as specified in the purchase order. Although the purchase order from Spectrum to WESCO specifically required new equipment and CoCs, WESCO purchased the MCCBs from a subvendor which dealt mainly in reconditioned equipment and provided these reconditioned MCCBs to Spectrum with falsified CoCs that certified that they were new equipment. In addition, the investigation identified that WESCO ordered Westinghouse labels from the subvendor in order to label the shipping boxes that lacked labels. Spectrum performed the dedication inspection and testing to demonstrate the adequacy of the MCCBs from WESCO. However, the validity of this testing depended on the MCCBs being new equipment. Spectrum's failure to verify the accuracy or the validity of the CoCs resulted in Spectrum accepting fraudulent CoCs and providing reconditioned (as opposed to new) MCCBs to Peach Bottom

  11. Westinghouse Savannah River Company: Report from the DOE Voluntary Protection Program onsite reviews, February 24--March 7, 1997, and June 15--19, 1998

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-05-01

    This report summarizes the Department of Energy Voluntary Protection Program (DOE-VPP) Initial and Update Review Teams` findings from the onsite evaluations of the Westinghouse Savannah River Site (SRS), conducted February 24--March 7, 1997, and June 15-19, 1998. The site was evaluated against the program requirements contained in US Department of Energy Voluntary Protection Program, Part 1: Program Elements to determine its success in implementing the five tenets of DOE-VPP. The Initial Review Team concluded that WSRC met or surpassed all DOE-VPP requirements, with the exception of 12 minor findings and 5 recommendations. WSRC was asked to resolve the findings within 90 days. During a follow-up visit in January 1996, representatives of the Team verified that all 90-day actions were completed. The Update Team detected though that the program did not demonstrate thorough and meaningful employee involvement. The ability to attain and sustain VPP-level performance on employee involvement is a significant challenge. Large companies with multiple layers of management and geographically disperse personnel have particular difficulty.

  12. Westinghouse Savannah River Company: Report from the DOE Voluntary Protection Program onsite reviews, February 24-March 7, 1997, and June 15-19, 1998

    International Nuclear Information System (INIS)

    1999-05-01

    This report summarizes the Department of Energy Voluntary Protection Program (DOE-VPP) Initial and Update Review Teams' findings from the onsite evaluations of the Westinghouse Savannah River Site (SRS), conducted February 24--March 7, 1997, and June 15-19, 1998. The site was evaluated against the program requirements contained in US Department of Energy Voluntary Protection Program, Part 1: Program Elements to determine its success in implementing the five tenets of DOE-VPP. The Initial Review Team concluded that WSRC met or surpassed all DOE-VPP requirements, with the exception of 12 minor findings and 5 recommendations. WSRC was asked to resolve the findings within 90 days. During a follow-up visit in January 1996, representatives of the Team verified that all 90-day actions were completed. The Update Team detected though that the program did not demonstrate thorough and meaningful employee involvement. The ability to attain and sustain VPP-level performance on employee involvement is a significant challenge. Large companies with multiple layers of management and geographically disperse personnel have particular difficulty

  13. Modular construction approach for advanced nuclear plants

    International Nuclear Information System (INIS)

    Johnson, F.T.; Orr, R.S.; Boudreaux, C.P.

    1988-01-01

    Modular construction has been designated as one of the major features of the AP600 program, a small innovative 600-MW (electric) advanced light water reactor (ALWR) that is currently being developed by Westinghouse and its subcontractors. This program is sponsored by the US Department of Energy (DOE) in conjunction with several other DOE and Electric Power Research Institute ALWR programs. Two major objectives of the AP600 program are as follows: (1) to provide a cost of power competitive with other power generation alternatives; and (2) to provide a short construction schedule that can be met with a high degree of certainty. The AP600 plant addresses these objectives by providing a simplified plant design and an optimized plant arrangement that result in a significant reduction in the number and size of systems and components, minimizes the overall building volumes, and consequently reduces the required bulk quantities. However, only by adopting a modular construction approach for the AP600 can the full cost and schedule benefits be realized from the advances made in the plant systems design and plant arrangement. Modularization is instrumental in achieving both of the above objectives, but most of all, a total modularization approach is considered absolutely essential to ensure that an aggressive construction schedule can be met with a high degree of certainty

  14. Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B. [and others

    1997-06-01

    This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked & influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs.

  15. Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

    International Nuclear Information System (INIS)

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B.

    1997-06-01

    This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked ampersand influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs

  16. Containment integrity analysis for the (W) advanced AP600

    International Nuclear Information System (INIS)

    Gagnon, A.F.; Howe, K.S.

    1989-01-01

    This paper reports that since 1987, Westinghouse has been performing containment cooling analyses in support of the Advanced AP600 plant design. This program was intended to verify the feasibility of the passive containment cooling system features of the AP600 design. To support this design, containment analyses of the AP600 containment for a large break LOCA and a large Steam Line Break were performed. The transient results indicate the feasibility of the passive containment design by demonstrating the capability to remove sufficient heat to limit containment atmosphere conditions to within acceptable limits following these postulated accidents. These results also indicate that the PCCS can reduce containment pressure to less than one-quarter design pressure at 24 hours following the most severe accident scenario thereby minimizing containment leakage concerns

  17. Advanced passive technology: A global standard for nuclear plant requirements

    Energy Technology Data Exchange (ETDEWEB)

    Novak, V

    1994-12-31

    Since 1984, Westinghouse has been developing AP8OO, a 800 MW, two-loop advanced passive plant, in response to an initiative established by the Electric Power Research Institute (EPRI) and the U.S. Department of Energy` (DOE). The preliminary design was cornpleved in 1989. AP6OO`s Standard Safety Analysis and Probabilistic Risk analysis Reports were submitted to the U.S. Nuclear Regulatory Commission for design certification in 1992. Design simplification is the key strategy behind the AP6OO. The basic technical concept Of simplification has resulted in a simplified reactor coolant systems, simplified plant systems, a simplified plant arrangement, reduced number of components, simplified operation and maintenance.

  18. US Advanced Light Water Reactor Program; overall objective

    International Nuclear Information System (INIS)

    Klug, N.

    1989-01-01

    The overall objective of the US Department of Energy (DOE) Advanced Light Water Reactor (ALWR) program is to perform coordinated programs of the nuclear industry and DOE to insure the availability of licensed, improved, and simplified light water reactor standard plant designs that may be ordered in the 1990's to help meet the US electrical power demand. The discussion includes plans to meet program objectives and the design certification program. DOE is currently supporting the development of conceptual designs, configurations, arrangements, construction methods/plans, and proof test key design features for the General Electric ASBWR and the Westinghouse AP600. Key features of each are summarized. Principal milestones related to licensing of large standard plants, simplified mid-size plant development, and plant lifetime improvement are noted

  19. Proceedings of the Advanced Turbine Systems annual program review meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-31

    Goals of the 8-year program are to develop cleaner, more efficient, and less expensive gas turbine systems for utility and industrial electric power generation, cogeneration, and mechanical drive units. During this Nov. 9-11, 1994, meeting, presentations on energy policy issues were delivered by representatives of regulatory, industry, and research institutions; program overviews and technical reviews were given by contractors; and ongoing and proposed future projects sponsored by university and industry were presented and displayed at the poster session. Panel discussions on distributed power and Advanced Gas Systems Research education provided a forum for interactive dialog and exchange of ideas. Exhibitors included US DOE, Solar Turbines, Westinghouse, Allison Engine Co., and GE.

  20. A survey on the development of advanced instrumentation and control system in NPP

    International Nuclear Information System (INIS)

    Ham, Chang Sik; Kwon, Kee Choon; Chung, Chul Hwan

    1993-12-01

    Many developed countries are improving or operating the advanced I and C systems of NPPs. They are: 1) N4 of EDF in France, 2) AP 600 of Westinghouse in USA, 3) NUPLEX-80+ of ABB-CE in USA, 4) CANDU in Canada, 5) Ohi 3 and 4, APWR and ABWR in Japan, 6) Belt-D in Germany, 7) Sizewell B in Britain, 8) Halden Reactor Projector in Norway, 9) I and C systems in Russia and Eastern Europe. This report describes the development trend, background, system architecture, characteristics with the new safety concerns, licensing problems, future plan, and retrofit experiences of these advanced nuclear I and C systems. The biggest difference between the existing systems and the advanced systems is the application of software rather than hardware for the functional implementation. All of the improved I and C systems accepted the standard modules and off-the shelf devices. Their characteristics are focused on EPRI URD Chapter 10. (author)

  1. A survey on the development of advanced instrumentation and control system in NPP

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Chang Sik; Kwon, Kee Choon; Chung, Chul Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-12-01

    Many developed countries are improving or operating the advanced I and C systems of NPPs. They are: (1) N4 of EDF in France, (2) AP 600 of Westinghouse in USA, (3) NUPLEX-80+ of ABB-CE in USA, (4) CANDU in Canada, (5) Ohi 3 and 4, APWR and ABWR in Japan, (6) Belt-D in Germany, (7) Sizewell B in Britain, (8) Halden Reactor Projector in Norway, (9) I and C systems in Russia and Eastern Europe. This report describes the development trend, background, system architecture, characteristics with the new safety concerns, licensing problems, future plan, and retrofit experiences of these advanced nuclear I and C systems. The biggest difference between the existing systems and the advanced systems is the application of software rather than hardware for the functional implementation. All of the improved I and C systems accepted the standard modules and off-the shelf devices. Their characteristics are focused on EPRI URD Chapter 10. (author).

  2. Advanced Ceramics

    International Nuclear Information System (INIS)

    1989-01-01

    The First Florida-Brazil Seminar on Materials and the Second State Meeting about new materials in Rio de Janeiro State show the specific technical contribution in advanced ceramic sector. The others main topics discussed for the development of the country are the advanced ceramic programs the market, the national technic-scientific capacitation, the advanced ceramic patents, etc. (C.G.C.) [pt

  3. Advanced light water reactors: an economically viable part of the world's future energy mix

    International Nuclear Information System (INIS)

    Bruschi, H.J.

    1996-01-01

    In addition to safety and reliability, a common mission for the international nuclear industry in the 21. century will be ensure affordable electricity. At the Westinghouse Electric Corporation believe our advanced light water reactor (ALWR) design gives us the opportunity to provide the safest, most reliable, lowest cost, most competitive generation method possible for use by nations and utilities worldwide. While the safety and reliability aspects of the ALWR can be proven tangibly and are well-documented, questions have been raised about the technology's ability to work within the world's selling price range for electricity generation. For our industry's financial stability, and especially for the stability of the world's future power needs, Westinghouse has done extensive work on this issue and we are convinced we can meet the competitive challenge. We believe the ALWR can be an economically viable part of the world's future energy mix. This paper will define the competitive challenge that is being addressed by the industry and then analyze three specific areas: capital costs, operating costs, and financing costs. The hidden advantage of nuclear power in responding to these challenges will be explored, and a strong case will be made asserting that the advanced light water reactor will be able to compete in international markets with viable production costs. (authors)

  4. 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.

  5. Human Performance Westinghouse Program; Programa Human Performance de Westinghouse

    Energy Technology Data Exchange (ETDEWEB)

    Garcia Gutierrez, A.; Gil, C.

    2010-07-01

    The objective of the Program consists in the excellence actuation, achieving the client success with a perfect realisation project. This program consists of different basic elements to reduce the human mistakes: the HuP tools, coaching, learning clocks and Know website. There is, too, a document file to consult and practice. All these elements are expounded in this paper.

  6. Westinghouse ICF power plant study

    International Nuclear Information System (INIS)

    Sucov, E.W.

    1980-10-01

    In this study, two different electric power plants for the production of about 1000 MWe which were based on a CO 2 laser driver and on a heavy ion driver have been developed and analyzed. The purposes of this study were: (1) to examine in a self consistent way the technological and institutional problems that need to be confronted and solved in order to produce commercially competitive electricity in the 2020 time frame from an inertial fusion reactor, and (2) to compare, on a common basis, the consequences of using two different drivers to initiate the DT fuel pellet explosions. Analytic descriptions of size/performance/cost relationships for each of the subsystems comprising the power plant have been combined into an overall computer code which models the entire plant. This overall model has been used to conduct trade studies which examine the consequences of varying critical design values around the reference point

  7. The IRIS consortium: international cooperation in advanced reactor development

    International Nuclear Information System (INIS)

    Carelli, M.; Petrovic, B.; Miller, K.; Lombardi, C.; Ricotti, M.E.

    2005-01-01

    after graduation, including hiring at Westinghouse. The innovative, advanced state of the art features of the IRIS design owe much to the world wide experience of its members and to the enthusiasm of a truly international cooperation with no boundaries. (authors)

  8. Advances in passive cooling design and performance analysis

    International Nuclear Information System (INIS)

    Woodcock, J.

    1994-01-01

    The Third International Conference on Containment Design and Operation continues the trend of rapidly extending the state of the art in containment methodology, joining other conferences, OECD-sponsored International Standard Problem exercises, and vendor licensing submittals. Methodology developed for use on plants with passive features is under increasing scrutiny for advanced designs, since the passive features are often the only deviation from existing operating base of the past 30 years of commercial nuclear power. This session, 'Containment Passive Safety Systems Design and Operation,' offers papers on a wide range of topics, with authors from six organizations from around the world, dealing with general passive containments, Westinghouse AP600, large (>1400 MWe) passive plants, and the AECL advanced CANDU reactor. This level and variety of participation underscores the high interest and accelerated methods development associated with advanced passive containment heat removal. The papers presented in this session demonstrate that significant contributions are being made to the advancement of technology necessary for building a new generation of safer, more economical nuclear plants. (author)

  9. Committed to the growth of the NP industry

    International Nuclear Information System (INIS)

    Agnihotri, Newal K.

    2004-01-01

    Mr. Stephen Tritch, President and Chief Executive Officer of Westinghouse Electric Company is responsible for al Westinghouse commercial nuclear operations, including the BNFL fuel business group in the United Kingdom. In this interview, he discusses economic aspects of bringing new power plants online, including waste disposal and investment issues. Also discussed are public relation activities to encourage public acceptance and what is needed from a practical and policy perspective to make new plant development happen in the U.S. Regarding the best available technology, he states that the AP1000 is the advanced nuclear power plant best-suited for new construction programs in the U.S. and elsewhere. It features passive and inherent safety systems, superior economics (3 to 3.5 cents per kilowatt hour) and modular design and construction that will help ensure highly predictable construction timetables. Lastly, Mr. Tritch discusses issues related to where the next generation of nuclear professionals will come from, including the knowledge transfer process, worldwide training standardization, utilizing retired professionals, and encouraging public schools to offer nuclear-based curriculum materials and intern programs

  10. Proceedings of the 2004 international congress on advances in nuclear power plants - ICAPP'04

    International Nuclear Information System (INIS)

    2004-01-01

    The 2004 International Congress on Advances in Nuclear Power Plants (ICAPP'04) provides a forum for the industry to exchange the latest ideas and research findings on nuclear plants from all perspectives. This conference builds on the success of last year's meeting held in Cordoba, Spain, and on the 2002 inaugural meeting held in Hollywood, Florida. Because of the hard work of many volunteers from around the world, ICAPP'04 has been successful in achieving its goal. More than 325 invited and contributed papers/presentations are part of this ICAPP. There are 5 invited plenary sessions and 70 technical sessions with contributed papers. The ICAPP'04 Proceedings contain almost 275 papers prepared by authors from 25 countries covering topics related to advances in nuclear power plant technology. The program by technical track deals with: 1 - Water-Cooled Reactor Programs and Issues (Status of All New Water-Cooled Reactor Programs; Advanced PWRs: Developmental Stage I; Advanced PWRs: Developmental Stage II; Advanced PWRs: Basic Design Stage; Advanced BWRs; Economics, Regulation, Licensing, and Construction; AP1000); 2 - High Temperature Gas Cooled Reactors (Pebble Bed Modular Reactors; Very High Temperature Reactors; HTR Fuels and Materials; Innovative HTRs and Fuel Cycles); 3 - Long Term Reactor Programs and Strategies (Supercritical Pressure Water Reactors; Lead-Alloy Fast Reactors; Sodium and Gas Fast Reactors; Status of Advanced Reactor Programs; Non-classical Reactor Concepts); 4 - Operation, Performance, and Reliability Management (Information Technology Effect on Plant Operation; Operation, Maintenance and Reliability; Improving Performance and Reducing O and M Costs; Plant Modernization and Retrofits); 5 - Plant Safety Assessment and Regulatory Issues (LOCA and non-LOCA Analysis Methodologies; LOCA and non-LOCA Plant Analyses; In-Vessel Retention; Containment Performance and Hydrogen Control; Advances in Severe Accident Analysis; Advances in Severe Accident

  11. Advanced fabrication technology

    International Nuclear Information System (INIS)

    Sheely, W.F.

    1986-01-01

    The Fuel Cycle Plant is a multipurpose nuclear facility located on the Hanford Nuclear Reservation in eastern Washington state. The facility is part of the Hanford Engineering Development Laboratory which is operated by Westinghouse Hanford Company for the Department of Energy. The Fuel Cycle Plant is currently being prepared to support the Liquid Metal Reactors Program with fuel fabrication services for the Fast Flux Test Facility and other LMR programs. This report describes the technical innovations to be utilized in the operation of this plant

  12. ADVANCE PAYMENTS

    CERN Multimedia

    Human Resources Division

    2002-01-01

    Administrative Circular Nº 8 makes provision for the granting of advance payments, repayable in several monthly instalments, by the Organization to the members of its personnel. Members of the personnel are reminded that these advances are only authorized in exceptional circumstances and at the discretion of the Director-General. In view of the current financial situation of the Organization, and in particular the loans it will have to incur, the Directorate has decided to restrict the granting of such advances to exceptional or unforeseen circumstances entailing heavy expenditure and more specifically those pertaining to social issues. Human Resources Division Tel. 73962

  13. Advance payments

    CERN Multimedia

    Human Resources Division

    2003-01-01

    Administrative Circular N 8 makes provision for the granting of advance payments, repayable in several monthly instalments, by the Organization to the members of its personnel. Members of the personnel are reminded that these advances are only authorized in exceptional circumstances and at the discretion of the Director-General. In view of the current financial situation of the Organization, and in particular the loans it will have to incur, the Directorate has decided to restrict the granting of such advances to exceptional or unforeseen circumstances entailing heavy expenditure and more specifically those pertaining to social issues. Human Resources Division Tel. 73962

  14. Implementation of an advanced digital feedwater control system at the Prairie Island nuclear generating station

    International Nuclear Information System (INIS)

    Paris, R.E.; Gaydos, K.A.; Hill, J.O.; Whitson, S.G.; Wirkkala, R.

    1990-05-01

    EPRI Project RP2126-4 was a cooperative effort between TVA, EPRI, and Westinghouse which resulted in the demonstration of a prototype of a full range, fully automatic feedwater control system, using fault tolerant digital technology, at the TVA Sequoyah simulator site. That prototype system also included advanced signal validation algorithms and an advanced man-machine interface that used CRT-based soft-control technology. The Westinghouse Advanced Digital Feedwater Control System (ADFCS) upgrade, which contains elements that were part of that prototype system, has since been installed at Northern States Power's Prairie Island Unit 2. This upgrade was very successful due to the use of an advanced control system design and the execution of a well coordinated joint effort between the utility and the supplier. The project experience is documented in this report to help utilities evaluate the technical implications of such a project. The design basis of the Prairie Island ADFCS signal validation for input signal failure fault tolerance is outlined first. Features of the industry-proven system control algorithms are then described. Pre-shipment hardware-in-loop and factory acceptance testing of the Prairie Island system are summarized. Post-shipment site testing, including preoperational and plant startup testing, is also summarized. Plant data from the initial system startup is included. The installation of the Prairie Island ADFCS is described, including both the feedwater control instrumentation and the control board interface. Modification of the plant simulator and operator and I ampersand C personnel training are also discussed. 6 refs., 14 figs., 3 tabs

  15. Advanced Electronics

    Science.gov (United States)

    2017-07-21

    AFRL-RV-PS- AFRL-RV-PS- TR-2017-0114 TR-2017-0114 ADVANCED ELECTRONICS Ashwani Sharma 21 Jul 2017 Interim Report APPROVED FOR PUBLIC RELEASE...NUMBER Advanced Electronics 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 62601F 6. AUTHOR(S) 5d. PROJECT NUMBER 4846 Ashwani Sharma 5e. TASK NUMBER...Approved for public release; distribution is unlimited. (RDMX-17-14919 dtd 20 Mar 2018) 13. SUPPLEMENTARY NOTES 14. ABSTRACT The Space Electronics

  16. AdvancED Flex 4

    CERN Document Server

    Tiwari, Shashank; Schulze, Charlie

    2010-01-01

    AdvancED Flex 4 makes advanced Flex 4 concepts and techniques easy. Ajax, RIA, Web 2.0, mashups, mobile applications, the most sophisticated web tools, and the coolest interactive web applications are all covered with practical, visually oriented recipes. * Completely updated for the new tools in Flex 4* Demonstrates how to use Flex 4 to create robust and scalable enterprise-grade Rich Internet Applications.* Teaches you to build high-performance web applications with interactivity that really engages your users.* What you'll learn Practiced beginners and intermediate users of Flex, especially

  17. Energy Conversion Alternatives Study (ECAS), Westinghouse phase 1. Volume 1: Introduction and summary and general assumptions. [energy conversion systems for electric power plants using coal - feasibility

    Science.gov (United States)

    Beecher, D. T.

    1976-01-01

    Nine advanced energy conversion concepts using coal or coal-derived fuels are summarized. They are; (1) open-cycle gas turbines, (2) combined gas-steam turbine cycles, (3) closed-cycle gas turbines, (4) metal vapor Rankine topping, (5) open-cycle MHD; (6) closed-cycle MHD; (7) liquid-metal MHD; (8) advanced steam; and (9) fuel cell systems. The economics, natural resource requirements, and performance criteria for the nine concepts are discussed.

  18. Assessment of EPRI water chemistry guidelines for new nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, K.; Fruzzetti, K.; Garcia, S. [Electric Power Research Inst., Palo Alto, California (United States); Eaker, R. [Richard W. Eaker, LLC, Matthews, North Carolina (United States); Giannelli, J.; Tangen, J. [Finetech, Inc., Parsippany, New Jersey (United States); Gorman, J.; Marks, C. [Dominion Engineering, Inc., Reston, Virginia (United States); Sawochka, S. [NWT Corp., San Jose, California (United States)

    2010-07-01

    Water chemistry control technologies for nuclear power plants have been significantly enhanced over the past few decades to improve material and equipment reliability and fuel performance, and to minimize radionuclide production and transport. Chemistry Guidelines have been developed by the Electric Power Research Institute (EPRI) for current operating plants and have been intermittently revised over the past twenty-five years for the protection of systems and components and for radiation management. As new plants are being designed for improved safety and increased power production, it is important to ensure that the designs consider implementation of industry approved water chemistry controls. In parallel, the industry will need to consider and develop updated water chemistry guidelines as well as plant startup and operational strategies based on the advanced plant designs. In 2010, EPRI began to assess chemistry control strategies at advanced plants, based on the Design Control Documents (DCDs), Combined Construction and Operating License Applications (COLA), and operating experiences (where they exist) against current Water Chemistry Guidelines. Based on this assessment, differences between planned chemistry operations at new plants and the current Guidelines will be identified. This assessment will form the basis of future activities to address these differences. The project will also assess and provide, as feasible, water chemistry guidance for startup and hot functional testing of the new plants. EPRI will initially assess the GE-Hitachi/Toshiba ABWR and the Westinghouse AP1000 designs. EPRI subsequently plans to assess other plant designs such as the AREVA U.S. EPR, Mitsubishi Heavy Industries (MHI) U.S. APWR, and GE-Hitachi (GE-H) ESBWR. This paper discusses the 2010 assessments of the ABWR and AP1000. (author)

  19. Assessment of EPRI water chemistry guidelines for new nuclear power plants

    International Nuclear Information System (INIS)

    Kim, K.; Fruzzetti, K.; Garcia, S.; Eaker, R.; Giannelli, J.; Tangen, J.; Gorman, J.; Marks, C.; Sawochka, S.

    2010-01-01

    Water chemistry control technologies for nuclear power plants have been significantly enhanced over the past few decades to improve material and equipment reliability and fuel performance, and to minimize radionuclide production and transport. Chemistry Guidelines have been developed by the Electric Power Research Institute (EPRI) for current operating plants and have been intermittently revised over the past twenty-five years for the protection of systems and components and for radiation management. As new plants are being designed for improved safety and increased power production, it is important to ensure that the designs consider implementation of industry approved water chemistry controls. In parallel, the industry will need to consider and develop updated water chemistry guidelines as well as plant startup and operational strategies based on the advanced plant designs. In 2010, EPRI began to assess chemistry control strategies at advanced plants, based on the Design Control Documents (DCDs), Combined Construction and Operating License Applications (COLA), and operating experiences (where they exist) against current Water Chemistry Guidelines. Based on this assessment, differences between planned chemistry operations at new plants and the current Guidelines will be identified. This assessment will form the basis of future activities to address these differences. The project will also assess and provide, as feasible, water chemistry guidance for startup and hot functional testing of the new plants. EPRI will initially assess the GE-Hitachi/Toshiba ABWR and the Westinghouse AP1000 designs. EPRI subsequently plans to assess other plant designs such as the AREVA U.S. EPR, Mitsubishi Heavy Industries (MHI) U.S. APWR, and GE-Hitachi (GE-H) ESBWR. This paper discusses the 2010 assessments of the ABWR and AP1000. (author)

  20. Fuel behavior in advanced water reactors

    International Nuclear Information System (INIS)

    Bolme, A.B.

    1996-01-01

    Fuel rod behavior of advanced pressurized water reactors under steady state conditions has been investigated in this study. System-80+ and Westinghouse Vantage-5 fuels have been considered as advanced pressurized water reactor fuels to be analyzed. The purpose of this study is to analyze the sensitivity of ditferent models and the effect of selected design parameters on the overall fuel behavior. FRAPCON-II computer code has been used for the analyses. Different modelling options of FRAPCON-II have also been considered in these analyses. Analyses have been performed in two main parts. In the first part, effects of operating conditions on fuel behavior have been investigated. First, fuel rod response under normal operating conditions has been analyzed. Then, fuel rod response to different fuel ratings has been calculated. In the second part, in order to estimate the effect of design parameters on fuel behavior, parametric analyses have been performed. In this part, the effects of initial gap thickness, as fabricated fuel density, and initial fill gas pressure on fuel behavior have been analyzed. The computations showed that both of the fuel rods used in this study operate within the safety limits. However, FRAPCON-II modelling options have been resulted in different behavior due to their modelling characteristics. Hence, with the absence of experimental data, it is difficult to make assesment for the best fuel parameters. It is also difficult to estimate error associated with the results. To improve the performance of the code, it is necessary to develop better experimental correlations for material properties in order to analyze the eftect ot considerably different design parameters rather than nominal rod parameters

  1. Advanced calculus

    CERN Document Server

    Nickerson, HK; Steenrod, NE

    2011-01-01

    ""This book is a radical departure from all previous concepts of advanced calculus,"" declared the Bulletin of the American Mathematics Society, ""and the nature of this departure merits serious study of the book by everyone interested in undergraduate education in mathematics."" Classroom-tested in a Princeton University honors course, it offers students a unified introduction to advanced calculus. Starting with an abstract treatment of vector spaces and linear transforms, the authors introduce a single basic derivative in an invariant form. All other derivatives - gradient, divergent, curl,

  2. Proposal for a advanced PWR core with adequate characteristics for passive safety concept

    International Nuclear Information System (INIS)

    Perrotta, Jose Augusto

    1999-01-01

    This work presents a discussion upon the suitable from an advanced PWR core, classified by the EPRI as 'Passive PWR' (advanced reactor with passive safety concept to power plants with less than 600 MW electrical power). The discussion upon the type of core is based on nuclear fuel engineering concepts. Discussion is made on type of fuel materials, structural materials, geometric shapes and manufacturing process that are suitable to produce fuel assemblies which give good performance for this type of reactors. The analysis is guided by the EPRI requirements for Advanced Light Water Reactor (ALWR). By means of comparison, the analysis were done to Angra 1 (old type of 600 MWe PWR class), and the design of the Westinghouse Advanced PWR-AP600. It was verified as a conclusion of this work that the modern PWR fuels are suitable for advanced PWR's Nevertheless, this work presents a technical alternative to this kind of fuel, still using UO 2 as fuel, but changing its cylindrical form of pellets and pin type fuel element to plane shape pallets and plate type fuel element. This is not a novelty fuel, since it was used in the 50's at Shippingport Reactor and as an advanced version by CEA of France in the 70's. In this work it is proposed a new mechanical assembly design for this fuel, which can give adequate safety and operational performance to the core of a 'Passive PWR'. (author)

  3. PENGEMBANGAN MODEL UNTUK SIMULASI KESELAMATAN REAKTOR PWR 1000 MWe GENERASI III+ MENGGUNAKAN PROGRAM KOMPUTER RELAP5

    Directory of Open Access Journals (Sweden)

    Andi Sofrany Ekariansyah

    2015-04-01

    Full Text Available Reaktor daya PWR AP1000 yang didesain oleh Westinghouse adalah reaktor Generasi III+ pertama yang telah menerima persetujuan desain dari U.S. Nuclear Regulatory Commission (NRC. Saat ini utilitas China telah memulai pembangunan beberapa unit AP1000 di dua tapak terpilih untuk rencana operasi pada 2013-2015. AP1000 sebagai desain PWR berdasarkan teknologi teruji dari desain PWR lainnya yang dibuat oleh Westinghouse dengan penguatan pada sistem keselamatan pasif dengan demikian dapat dipertimbangkan untuk dibangun di Indonesia bila mengacu pada persyaratan pada PP 43/2006 mengenai Perijinan Reaktor Nuklir. Namun demikian, desain tersebut perlu diverifikasi oleh Technical Support Organization (TSO independen sebelum dapat dibangun di Indonesia. Verifikasi dapat dilakukan menggunakan paket program RELAP5 dalam bentuk analisis kecelakaan. Selama ini analisis kecelakaan PLTN dilakukan untuk tipe PWR 1000 MWe dari generasi II atau tipe konvensional. Mengingat saat ini referensi yang menggambarkan teknologi AP1000 yang menyertakan teknologi keselamatan pasif sudah tersedia maka dilakukan kegiatan pemodelan yang nantinya dapat digunakan untuk melakukan analisis kecelakaan. Metode pengembangan model mengacu pada pedoman IAEA yang terdiri dari pengumpulan data instalasi, pengembangan engineering data dan penyusunan input deck, verifikasi dan validasi data input. Model yang berhasil dikembangkan secara umum telah mewakili sistem AP1000 secara keseluruhan dan dianggap sebagai model dasar. Model tersebut telah diverifikasi dan divalidasi dengan data desain yang terdapat pada referensi dimana respon parameter termohidraulika menunjukkan perbedaan hasil ± 3% selain untuk parameter penurunan tekanan teras yang lebih rendah 13%. Sebagai model dasar, input deck yang diperoleh dapat dikembangkan lebih lanjut dengan mengintegrasikan pemodelan sistem keselamatan, sistem proteksi, dan sistem kendali yang spesifik AP1000 untuk keperluan simulasi keselamatan yang lebih

  4. Advanced Virgo

    CERN Multimedia

    Virgo, a first-generation interferometric gravitational wave (GW) detector, located in the European Gravitational Observatory, EGO, Cascina (Pisa-Italy) and constructed by the collaboration of French and Italian institutes (CNRS and INFN) has successfully completed its long-duration data taking runs. It is now undergoing a fundamental upgrade that exploits available cutting edges technology to open an exciting new window on the universe, with the first detection of a gravitational wave signal. Advanced Virgo (AdV) is the project to upgrade the Virgo detector to a second-generation instrument. AdV will be able to scan a volume of the Universe 1000 times larger than initial Virgo. AdV will be hosted in the same infrastructures as Virgo. The Advanced VIRGO project is funded and at present carried on by a larger collaboration of institutes belonging to CNRS- France , RMKI - Hungary, INFN- Italy, Nikhef - The Netherlands Polish Academy of Science - Poland.

  5. Advanced Combustion

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, Gordon R. [NETL

    2013-03-11

    The activity reported in this presentation is to provide the mechanical and physical property information needed to allow rational design, development and/or choice of alloys, manufacturing approaches, and environmental exposure and component life models to enable oxy-fuel combustion boilers to operate at Ultra-Supercritical (up to 650{degrees}C & between 22-30 MPa) and/or Advanced Ultra-Supercritical conditions (760{degrees}C & 35 MPa).

  6. Future Transient Testing of Advanced Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Jon Carmack

    2009-09-01

    The transient in-reactor fuels testing workshop was held on May 4–5, 2009 at Idaho National Laboratory. The purpose of this meeting was to provide a forum where technical experts in transient testing of nuclear fuels could meet directly with technical instrumentation experts and nuclear fuel modeling and simulation experts to discuss needed advancements in transient testing to support a basic understanding of nuclear fuel behavior under off-normal conditions. The workshop was attended by representatives from Commissariat à l'Énergie Atomique CEA, Japanese Atomic Energy Agency (JAEA), Department of Energy (DOE), AREVA, General Electric – Global Nuclear Fuels (GE-GNF), Westinghouse, Electric Power Research Institute (EPRI), universities, and several DOE national laboratories. Transient testing of fuels and materials generates information required for advanced fuels in future nuclear power plants. Future nuclear power plants will rely heavily on advanced computer modeling and simulation that describes fuel behavior under off-normal conditions. TREAT is an ideal facility for this testing because of its flexibility, proven operation and material condition. The opportunity exists to develop advanced instrumentation and data collection that can support modeling and simulation needs much better than was possible in the past. In order to take advantage of these opportunities, test programs must be carefully designed to yield basic information to support modeling before conducting integral performance tests. An early start of TREAT and operation at low power would provide significant dividends in training, development of instrumentation, and checkout of reactor systems. Early start of TREAT (2015) is needed to support the requirements of potential users of TREAT and include the testing of full length fuel irradiated in the FFTF reactor. The capabilities provided by TREAT are needed for the development of nuclear power and the following benefits will be realized by

  7. Future Transient Testing of Advanced Fuels

    International Nuclear Information System (INIS)

    Carmack, Jon

    2009-01-01

    The transient in-reactor fuels testing workshop was held on May 4-5, 2009 at Idaho National Laboratory. The purpose of this meeting was to provide a forum where technical experts in transient testing of nuclear fuels could meet directly with technical instrumentation experts and nuclear fuel modeling and simulation experts to discuss needed advancements in transient testing to support a basic understanding of nuclear fuel behavior under off-normal conditions. The workshop was attended by representatives from Commissariat energie Atomique CEA, Japanese Atomic Energy Agency (JAEA), Department of Energy (DOE), AREVA, General Electric - Global Nuclear Fuels (GE-GNF), Westinghouse, Electric Power Research Institute (EPRI), universities, and several DOE national laboratories. Transient testing of fuels and materials generates information required for advanced fuels in future nuclear power plants. Future nuclear power plants will rely heavily on advanced computer modeling and simulation that describes fuel behavior under off-normal conditions. TREAT is an ideal facility for this testing because of its flexibility, proven operation and material condition. The opportunity exists to develop advanced instrumentation and data collection that can support modeling and simulation needs much better than was possible in the past. In order to take advantage of these opportunities, test programs must be carefully designed to yield basic information to support modeling before conducting integral performance tests. An early start of TREAT and operation at low power would provide significant dividends in training, development of instrumentation, and checkout of reactor systems. Early start of TREAT (2015) is needed to support the requirements of potential users of TREAT and include the testing of full length fuel irradiated in the FFTF reactor. The capabilities provided by TREAT are needed for the development of nuclear power and the following benefits will be realized by the

  8. Small break loss of coolant accident analysis of advanced PWR plant designs utilizing DVI line venturis

    International Nuclear Information System (INIS)

    Kemper, Robert M.; Gagnon, Andre F.; McNamee, Kevin; Cheung, Augustine C.

    1995-01-01

    The Westinghouse Advanced Passive and evolutionary Pressurizer Water Reactors (i.e. AP600 and APWR) incorporate direct vessel injection (DVI) of emergency core coolant as a means of minimizing the potential spilling of emergency core cooling water during a loss of coolant accident (LOCA). As a result, the most limiting small break LOCA (SBLOCA) event for these designs, with respect core inventory makeup capability, is a postulated double ended rupture of one of the DVI lines. This paper presents the results of a design optimization study that examines the installation of a venturi in the DVI line as a means of limiting the reactor coolant lost from the reactor vessel. The comparison results demonstrate that by incorporating a properly sized venturi in the DVI line, core uncovery concerns as a result of a DVI line break can be eliminated for both the AP600 and APWR plants. (author)

  9. Irradiation performance updates on Korean advanced fuels for PWRs

    International Nuclear Information System (INIS)

    Jang, Y.K.; Jeon, K.L.; Kim, Y.H.; Yoo, J.S.; Kim, J.I.; Shin, J.C.; Chung, J.G.; Park, J.R.; Chung, S.K.; Kim, T.W.; Yoon, Y.B.; Park, K.M.; Yoo, M.J.; Kim, M.S.; Lee, T.H.

    2010-01-01

    The developments of advanced nuclear fuels for PWRs were started in 1999 and in 2001, respectively: PLUS7 TM for eight operating optimized power reactors of 1000 MWe class (OPR1000) and four advanced power reactors of 1400 MWe class (APR1400) under construction, and 16ACE7 TM and 17ACE7 TM for an operating 16x16 Westinghouse type plant and six operating 17x17 Westinghouse type plants. The design targets were as follows: batch average burnup up to 55 GWD/MTU, over 10% thermal margin increase, improvement of the mechanical integrity of higher seismic capability, higher debris or grid fretting wear performance, higher control rod insertion capability, increase of neutron economy, improvement of manufacturability, solving incomplete rod insertion (IRI) issue and top nozzle screw failure issue, etc. in comparison of the existing nuclear fuels. The irradiation tests using each four LTAs (Lead Test Assemblies) during 3 cycles were completed in three Korean nuclear reactors until 2009. The eight irradiation performance items which are assembly growth, rod growth, grid width growth, assembly bow, rod bow, assembly twist, rod diameter and cladding oxidation were examined in pool-side after each cycle and evaluated. The irradiation tests could be continued by expecting the good performances for next cycle from the previous cycle. After 2 cycle irradiations, the region implementation could be started in 15 nuclear power plants. Even though the verifications using the LTAs were completed, each surveillance program was launched and the irradiation performance data were being updated during region implementation. In addition to pool-side examinations (PSEs) by assembly-wise during irradiation tests, six rod-wise performance items were also examined in pool-side using each LTA after discharge. All performance items met their design criteria as a result of the evaluation. Even though the interesting ones among the irradiation performance parameters were assembly and grid growths

  10. MSLB coupled 3D neutronics-thermalhydraulic analysis of a large PWR using RELAP5-3D

    International Nuclear Information System (INIS)

    Lo Nigro, A.; Spadoni, A.; D'Auria, F.; Saiu, G.

    2001-01-01

    A RELAP5-3D model of the Westinghouse AP1000 NSSS has been set up and it has been used to analyze the MSLB accident. Main results (both spatial distributions and time trends) have been represented with 3D plots and graphical movies. The method applied allows accounting for the coupled 3D neutronics and thermalyhdraulics effects, suggesting to consider its applicability in Safety Analysis.(author)

  11. Latest development in project site radwaste treatment facility (SRTF) Sanmen

    International Nuclear Information System (INIS)

    Mennicken, K.; Lohmann, P.

    2015-01-01

    Westinghouse Electric Germany GmbH (WEG) was successful in being awarded a contract as to the planning, delivery, installation and commissioning of radwaste treatment systems for the AP1000 units at Sanmen site, PR China. Operational low and intermediate level radioactive waste will be processed in the Site Radwaste Treatment Facility (SRTF). This paper explains the latest developments of the project, especially the experience with customer-hired Chinese planning partners, installation companies and Customer operating personnel. (authors)

  12. Advanced calculus

    CERN Document Server

    Friedman, Avner

    2007-01-01

    This rigorous two-part treatment advances from functions of one variable to those of several variables. Intended for students who have already completed a one-year course in elementary calculus, it defers the introduction of functions of several variables for as long as possible, and adds clarity and simplicity by avoiding a mixture of heuristic and rigorous arguments.The first part explores functions of one variable, including numbers and sequences, continuous functions, differentiable functions, integration, and sequences and series of functions. The second part examines functions of several

  13. Advanced calculus

    CERN Document Server

    Fitzpatrick, Patrick M

    2009-01-01

    Advanced Calculus is intended as a text for courses that furnish the backbone of the student's undergraduate education in mathematical analysis. The goal is to rigorously present the fundamental concepts within the context of illuminating examples and stimulating exercises. This book is self-contained and starts with the creation of basic tools using the completeness axiom. The continuity, differentiability, integrability, and power series representation properties of functions of a single variable are established. The next few chapters describe the topological and metric properties of Euclide

  14. Advanced trigonometry

    CERN Document Server

    Durell, C V; Robson, A

    1950-01-01

    This volume will provide a welcome resource for teachers seeking an undergraduate text on advanced trigonometry, when few are readily available. Ideal for self-study, this text offers a clear, logical presentation of topics and an extensive selection of problems with answers. Contents include the properties of the triangle and the quadrilateral; equations, sub-multiple angles, and inverse functions; hyperbolic, logarithmic, and exponential functions; and expansions in power-series. Further topics encompass the special hyperbolic functions; projection and finite series; complex numbers; de Moiv

  15. Meeting India's growing energy demand with nuclear power

    International Nuclear Information System (INIS)

    Matzie, R.

    2009-01-01

    Full text: With world energy demand expected to nearly double by 2030, the need for safe, reliable and clean energy is imperative. In India, energy demand has outpaced the increase in energy production, with the country experiencing as much as a 12 percent gap between peak demand and availability. To meet demand, nuclear power is the ideal solution for providing baseload electricity, and as much as 40-60 GWe of nuclear capacity will need to be added throughout the county over the next 20 years. This presentation will describe the benefits of nuclear power compared to other energy sources, provide an overview of new nuclear power plant construction projects worldwide, and explain the benefits and advantages of the Westinghouse AP1000 nuclear power plant. The presentation will also outline the steps that Westinghouse is taking to help facilitate new nuclear construction in India, and how the company's 'Buy Where We Build' approach to supply chain management will positively impact the Indian economy through continued in-country supplier agreements, job creation, and the exporting of materials and components to support AP1000 projects outside of India. Finally, the presentation will show that the experience Westinghouse is gaining in constructing AP1000 plants in both China and the United States will help ensure the success of projects in India

  16. Advanced Pacemaker

    Science.gov (United States)

    1990-01-01

    Synchrony, developed by St. Jude Medical's Cardiac Rhythm Management Division (formerly known as Pacesetter Systems, Inc.) is an advanced state-of-the-art implantable pacemaker that closely matches the natural rhythm of the heart. The companion element of the Synchrony Pacemaker System is the Programmer Analyzer APS-II which allows a doctor to reprogram and fine tune the pacemaker to each user's special requirements without surgery. The two-way communications capability that allows the physician to instruct and query the pacemaker is accomplished by bidirectional telemetry. APS-II features 28 pacing functions and thousands of programming combinations to accommodate diverse lifestyles. Microprocessor unit also records and stores pertinent patient data up to a year.

  17. The Nuclear Renaissance — Implications on Quantitative Nondestructive Evaluations

    Science.gov (United States)

    Matzie, Regis A.

    2007-03-01

    The world demand for energy is growing rapidly, particularly in developing countries that are trying to raise the standard of living for billions of people, many of whom do not even have access to electricity. With this increased energy demand and the high and volatile price of fossil fuels, nuclear energy is experiencing resurgence. This so-called nuclear renaissance is broad based, reaching across Asia, the United States, Europe, as well as selected countries in Africa and South America. Some countries, such as Italy, that have actually turned away from nuclear energy are reconsidering the advisability of this design. This renaissance provides the opportunity to deploy more advanced reactor designs that are operating today, with improved safety, economy, and operations. In this keynote address, I will briefly present three such advanced reactor designs in whose development Westinghouse is participating. These designs include the advanced passive PWR, AP1000, which recently received design certification for the US Nuclear Regulatory Commission; the Pebble Bed Modular reactor (PBMR) which is being demonstrated in South Africa; and the International Reactor Innovative and Secure (IRIS), which was showcased in the US Department of Energy's recently announced Global Nuclear Energy Partnership (GNEP), program. The salient features of these designs that impact future requirements on quantitative nondestructive evaluations will be discussed. Such features as reactor vessel materials, operating temperature regimes, and new geometric configurations will be described, and mention will be made of the impact on quantitative nondestructive evaluation (NDE) approaches.

  18. The Nuclear Renaissance - Implications on Quantitative Nondestructive Evaluations

    International Nuclear Information System (INIS)

    Matzie, Regis A.

    2007-01-01

    The world demand for energy is growing rapidly, particularly in developing countries that are trying to raise the standard of living for billions of people, many of whom do not even have access to electricity. With this increased energy demand and the high and volatile price of fossil fuels, nuclear energy is experiencing resurgence. This so-called nuclear renaissance is broad based, reaching across Asia, the United States, Europe, as well as selected countries in Africa and South America. Some countries, such as Italy, that have actually turned away from nuclear energy are reconsidering the advisability of this design. This renaissance provides the opportunity to deploy more advanced reactor designs that are operating today, with improved safety, economy, and operations. In this keynote address, I will briefly present three such advanced reactor designs in whose development Westinghouse is participating. These designs include the advanced passive PWR, AP1000, which recently received design certification for the US Nuclear Regulatory Commission; the Pebble Bed Modular reactor (PBMR) which is being demonstrated in South Africa; and the International Reactor Innovative and Secure (IRIS), which was showcased in the US Department of Energy's recently announced Global Nuclear Energy Partnership (GNEP), program. The salient features of these designs that impact future requirements on quantitative nondestructive evaluations will be discussed. Such features as reactor vessel materials, operating temperature regimes, and new geometric configurations will be described, and mention will be made of the impact on quantitative nondestructive evaluation (NDE) approaches

  19. Commercializing the next generation: the AP600 advanced simplified nuclear power plant

    International Nuclear Information System (INIS)

    Bruschi, H.J.

    1994-01-01

    Today, government and industry are working together on advanced nuclear power plant designs that take advantage of valuable lessons learned from the experience to date and promise to reconcile the demands of economic expansion with the laws of environmental protection. In the U.S., the Department of Energy (DOE) and the Electric Power Research Institute (EPRI) initiated a design certification program in 1989 to develop and commercialize advanced light water reactors (ALWRs) for the next round of power plant construction. Advanced, simplified technology is one approach under development to end the industry's search for a simpler, more forgiving, and less costly reactor. As part of this program, Westinghouse is developing the AP600, a new standard 600 MWe advanced, simplified plant. The design strikes a balance between the use of proven technology and new approaches. The result is a greatly streamlined plant that can meet safety regulations and reliability requirements, be economically competitive, and promote broader public confidence in nuclear energy. 1 fig

  20. PLUS 7TM advanced fuel assembly development program for KSNPs and APR1400

    International Nuclear Information System (INIS)

    Kim, Kyutae; Stucker, David L.

    2002-01-01

    KNFC and Westinghouse have recently completed the development of the PLUS 7 TM advanced 16 X 16 fuel assembly for the Korean Standard Nuclear Plants (KSNPs) and the Advanced Power Reactor 1400 (APR 1400). This fuel design utilized the proven advanced design features including mixing vane spacer grids to increase critical heat flux performance, ZIRLO TM advanced materials to enable high-duty, high burnup fuel management and an optimized fuel rod diameter which improves fuel cycle cost while resulting in significant standardization of Korean fuel manufacture. PLUS 7 TM , also includes a patented spacer grid design with conformal fuel rod support designed to provide superior fuel rod wear/fretting resistance while minimizing pressure drop. This paper will present an overview of the PLUS 7 TM fuel assembly development process including a summary of the three-year design and testing program from a mechanical, neutronic, and thermal/hydraulic perspective. The PLUS 7 TM fuel for the KSNPs and the APR1400 reactors results in multi-million dollar per cycle savings in imported enriched uranium product for the Korean nuclear power program with technology specifically developed for Korea by experienced Korean engineers