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Sample records for westinghouse pressurized water

  1. 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

  2. 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)

  3. 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

  4. 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

  5. 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

  6. 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-pressure

  7. 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

  8. 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)

  9. 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)

  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. 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.

  12. 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.

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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)

  19. 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

  20. 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

  1. 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...

  2. 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

  3. 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)

  4. 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)

  5. 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

  6. 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)

  7. 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

  8. 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

  9. RELAP5 thermal-hydraulic analyses of overcooling sequences in a pressurized water reactor

    International Nuclear Information System (INIS)

    Bolander, M.A.; Fletcher, C.D.; Davis, C.B.; Kullberg, C.M.; Stitt, B.D.; Waterman, M.E.; Burtt, J.D.

    1984-01-01

    In support of the Pressurized Thermal Shock Integration Study, sponsored by the United States Nuclear Regulatory Commission, the Idaho National Engineering Laboratory has performed analyses of overcooling transients using the RELAP5/MOD1.6 and MOD2.0 computer codes. These analyses were performed for the H.B. Robinson Unit 2 pressurized water reactor, which is a Westinghouse 3-loop design plant. Results of the RELAP5 analyses are presented. The capabilities of the RELAP5 computer code as a tool for analyzing integral plant transients requiring a detailed plant model, including complex trip logic and major control systems, are examined

  10. 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

  11. Modeling of hydrogen stratification in a pressurized water reactor containment with the contain computer code

    International Nuclear Information System (INIS)

    Kljenak, I.; Skerlavaj, A.; Parzer, I.

    1999-01-01

    Hydrogen distribution during a severe accident in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN computer code. The accidents is initiated by a large-break loss-of-coolant accident which is nit successfully mitigated by the action of the emergency core cooling system. Cases with and without successful actuation of spray systems and fan coolers were considered. The simulations predicted hydrogen stratification within the containment main compartment with intensive hydrogen mixing in the containment dome region. Pressure and temperature responses were analyzed as well.(author)

  12. 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

  13. Moderator feedback effects in two-dimensional nodal methods for pressurized water reactor analysis

    International Nuclear Information System (INIS)

    Downar, T.J.

    1987-01-01

    A method was developed for incorporating moderator feedback effects in two-dimensional nodal codes used for pressurized water reactor (PWR) neutronic analysis. Equations for the assembly average quality and density are developed in terms of the assembly power calculated in two dimensions. The method is validated with a Westinghouse PWR using the Electric Power Research Institute code SIMULATE-E. Results show a several percent improvement is achieved in the two-dimensional power distribution prediction compared to methods without moderator feedback

  14. 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)

  15. 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)

  16. 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

  17. 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

  18. Effect of Spray System on Fission Product Distribution in Containment During a Severe Accident in a Two-Loop Pressurized Water Reactor

    Directory of Open Access Journals (Sweden)

    Mehdi Dehjourian

    2016-08-01

    Full Text Available The containment response during the first 24 hours of a low-pressure severe accident scenario in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN 2.0 computer code. The accident considered in this study is a large-break loss-of-coolant accident, which is not successfully mitigated by the action of safety systems. The analysis includes pressure and temperature responses, as well as investigation into the influence of spray on the retention of fission products and the prevention of hydrogen combustion in the containment.

  19. 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.)

  20. 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

  1. 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.

  2. 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

  3. 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)

  4. 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)

  5. A feasibility study on feed and bleed for pressurized water reactors

    International Nuclear Information System (INIS)

    Yi-Shung Chen; Shimeck, D.J.; Sullivan, L.H.

    1983-01-01

    By injecting coolant with a high pressure emergency core cooling system, and removing the heated/ vaporized fluid by way of the pressurizer power operated relief valve, primary feed and bleed cooling denotes an operation whereby reactor core cooling is maintained. This paper presents the results from an experimental and analytical study that includes a simplified analysis of mass and energy balances associated with the feed and bleed, examination of test data from the Semiscale system, RELAP5 code analyses of both Semiscale and a four-loop Westinghouse plant, and the primary coolant system behavior for a transient that leads to the need for feed and bleed. Examination of the parameters that govern a stable feed and bleed operation identifies four key parameters such as: core decay heat, cooling water injection capacity, power operated relief valve (PORV) energy removal rate, and PORV mass removal rate. A simplified analytical approach to determining if stable feed and bleed is feasible, has been developed and corroborated by experimental data and computer code calculations. The Semiscale tests have not only provided test data for code assessment, but also have identified the factors influencing the PORV discharge, which is the most variable of the boundary conditions influencing feed and bleed. The RELAP5 computer code has demonstrated the capability of predicting the Semiscale experiments, and when applied to a four-loop Westinghouse plant has indicated that primary feed and bleed is a viable cooling mechanism. This has also been shown by using the simplified analytical method

  6. Study on severe accident induced by large break loss of coolant accident for pressureized water reactor

    International Nuclear Information System (INIS)

    Zhang Longfei; Zhang Dafa; Wang Shaoming

    2007-01-01

    Using the best estimate computer code SCDAP/RELAP5/MOD3.2 and taking US Westinghouse corporation Surry nuclear power plant as the reference object, a typical three-loop pressurized water reactor severe accident calculation model was established and 25 cm large break loss of coolant accident (LBLOCA) in cold and hot leg of primary loop induced core melt accident was analyzed. The calculated results show that core melt progression is fast and most of the core material melt and relocated to the lower plenum. The lower head of reactor pressure vessel failed at an early time and the cold leg break is more severe than the hot leg break in primary loop during LBLOCA. (authors)

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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)

  12. 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

  13. 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

  14. 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

  15. Contemporary pressurized water reactor technology in the world

    International Nuclear Information System (INIS)

    Komarek, A.

    1991-01-01

    The recent political events enabled Czechoslovak industrial companies to come into direct contact with leading western companies involved in pressurized water ractor technology. A survey is presented of the present situation at the world market of PWR type nuclear power plant suppliers and suppliers of fuel cycle services. Information is given on the potential bids for the next Czechoslovak nuclear power plants with PWR reactors. Economic aspects of the potential bids are presented including some considerations about the participation of the Czechoslovak nuclear industry as a supplier of the reactor for the future power plants. Main technical parameters are listed of PWR units with an output about 1000 MW supplied by Westinghouse EC, ABB -Combustion Engineering and Siemens AG. At present, the bids for new Czechoslovak nuclear power plants are being evaluated. No information on terms of the bids actually coming from foreign companies is used in the article. (Z.S.). 9 figs., 5 tabs

  16. Mitigation of stress corrosion cracking in pressurized water reactor (PWR) piping systems using the mechanical stress improvement process (MSIPR) or underwater laser beam welding

    International Nuclear Information System (INIS)

    Rick, Grendys; Marc, Piccolino; Cunthia, Pezze; Badlani, Manu

    2009-01-01

    A current issue facing pressurized water reactors (PWRs) is primary water stress corrosion cracking (PWSCC) of bi metallic welds. PWSCC in a PWR requires the presence of a susceptible material, an aggressive environment and a tensile stress of significant magnitude. Reducing the potential for SCC can be accomplished by eliminating any of these three elements. In the U.S., mitigation of susceptible material in the pressurizer nozzle locations has largely been completed via the structural weld overlay (SWOL) process or NuVision Engineering's Mechanical Stress Improvement Process (MSIP R) , depending on inspectability. The next most susceptible locations in Westinghouse designed power plants are the Reactor Vessel (RV) hot leg nozzle welds. However, a full SWOL Process for RV nozzles is time consuming and has a high likelihood of in process weld repairs. Therefore, Westinghouse provides two distinctive methods to mitigate susceptible material for the RV nozzle locations depending on nozzle access and utility preference. These methods are the MSIP and the Underwater Laser Beam Welding (ULBW) process. MSIP applies a load to the outside diameter of the pipe adjacent to the weld, imposing plastic strains during compression that are not reversed after unloading, thus eliminating the tensile stress component of SCC. Recently, Westinghouse and NuVision successfully applied MSIP on all eight RV nozzles at the Salem Unit 1 power plant. Another option to mitigate SCC in RV nozzles is to place a barrier between the susceptible material and the aggressive environment. The ULBW process applies a weld inlay onto the inside pipe diameter. The deposited weld metal (Alloy 52M) is resistant to PWSCC and acts as a barrier to prevent primary water from contacting the susceptible material. This paper provides information on the approval and acceptance bases for MSIP, its recent application on RV nozzles and an update on ULBW development

  17. 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

  18. 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.

  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. 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

  1. Preliminary evaluation of SACI-O code for the analysis of transients in a pressurized water reactor core

    International Nuclear Information System (INIS)

    Soares, P.A.; Sirimarco, L.F.; Veloso, M.A.F.

    1979-03-01

    SACI-O is a computer code which deals with the dynamics of the core of pressurized light water reactors (PWR). Its applicability is determined by the evaluation of the models used in the simulation of the several phenomena and processes which occur in the core during transients. This report presents a comparison between the results obtained with SACI-O and those presented in the Final Safety Analysis Report (FSAR) of Angra dos Reis Nuclear Station, Unit 1. Although some data used in the calculations done by Westinghouse are not known, there was a good agreement between the mentioned results. (Author) [pt

  2. 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

  3. 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.)

  4. Pressure behaviour in a nuclear reactor containment following a loss of coolant accident

    International Nuclear Information System (INIS)

    KHattab, M.S.; Ibrahim, N.A.; Bedrose, S.D.

    1994-01-01

    The scenarios of pressure variation following a loss of coolant accident (LOCA) inside the containment of pressurized water reactor (PWR) have been investigated. Critical mass flow rushing out from high pressure leg through pipe break, is used to calculate the rate of coolant. The energy added to the containment atmosphere is determined to specify the rate of growth of pressure and temperature. The scenarios of small, medium and large LOCA at 2%, 15% and 25% flow released are investigated. Safety water spray system is initiated as the pressure reaches the containment design safety limit at about 3 bar to depressurise and to cooldown the system and thereby to reduce the concentration of radioactivity release in the containment atmosphere. The pressure response before and after operation of safety spray system is predicted in each size of LOCA using a typical design of westinghouse PWR system. The results of large LOCA showed good agreement with westinghouse calculations of the same design. The heat removal from the containment environment is rejected into the sump by drop-wise condensation mechanism. The effect of initial droplets diameters injected from the nozzles of the spray system is investigated. The results show that the droplet diameter of 3 mm gives best performance. 6 figs., 1 tab

  5. 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

  6. 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

  7. 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)

  8. 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

  9. An estimation of core damage frequency of a pressurized water reactor during mid-loop operation

    International Nuclear Information System (INIS)

    Chao, C.C.; Chen, C.T.; Lee, M.

    2004-01-01

    The core damage frequency during mid-loop operation of a Westinghouse designed 3-loop Pressurizer Water Reactor (PWR) due to loss of Residual Heat Removal (RHR) events was assessed. The assessment considers two types of outages (refueling and drained maintenance), and uses failure data collected specifically for shutdown condition. Event trees were developed for five categories of loss of RHR events. Human actions to mitigate the loss of RHR events was identified and human error probabilities were quantified using HCR and THERP model. The result showed that the core damage frequency due to loss of RHR events during mid-loop operation is 3.1x10 -5 per year. The results also showed that the core damage frequency can be reduced significantly by removing a pressurizer safety valve before entering mid-loop operation. The establishment of reflux cooling, i.e. decay heat removal through steam generator secondary side also plays important role in mitigating the loss of RHR events. (author)

  10. 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

  11. 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

  12. 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.

  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. 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

  15. 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

  16. 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

  17. 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)

  18. 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)

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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)

  4. 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

  5. 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)

  6. Taipower's reload safety evaluation methodology for pressurized water reactors

    International Nuclear Information System (INIS)

    Huang, Ping-Hue; Yang, Y.S.

    1996-01-01

    For Westinghouse pressurized water reactors (PWRs) such as Taiwan Power Company's (TPC's) Maanshan Units 1 and 2, each of the safety analysis is performed with conservative reload related parameters such that reanalysis is not expected for all subsequent cycles. For each reload cycle design, it is required to perform a reload safety evaluation (RSE) to confirm the validity of the existing safety analysis for fuel cycle changes. The TPC's reload safety evaluation methodology for PWRs is based on 'Core Design and Safety Analysis Package' developed by the TPC and the Institute of Nuclear Energy Research (INER), and is an important portion of the 'Taipower's Reload Design and Transient Analysis Methodologies for Light Water Reactors'. The Core Management System (CMS) developed by Studsvik of America, the one-dimensional code AXINER developed by TPC, National Tsinghua University and INER, and a modified version of the well-known subchannel core thermal-hydraulic code COBRAIIIC are the major computer codes utilized. Each of the computer models is extensively validated by comparing with measured data and/or vendor's calculational results. Moreover, parallel calculations have been performed for two Maanshan reload cycles to validate the RSE methods. The TPC's in-house RSE tools have been applied to resolve many important plant operational issues and plant improvements, as well as to verify the vendor's fuel and core design data. (author)

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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

    plants dismantled to date in the US have repackaged the less activated waste back into the reactor vessel and shipped the entire assembly to the disposal site. Decisions like these can be driven by many factors such as disposal costs, transportation logistics, licensing fees, etc., but will have a significant impact on the segmentation and packaging plan so must be considered early in the planning phase. All segmentation tools are remotely controlled since the mechanical segmentation projects that Westinghouse has executed, so far, have been performed under water due to the high radiation levels. ALARA and personal safety is the number one priority during the site work. The complexity of the work requires well designed and reliable tools. Westinghouse has optimized the technologies from its experiences accumulated over the years. Its main focus has always been to improve tool handling and cutting speed, water cleanliness, fail-safe and safety aspects. Different band saws, disc saws, tube cutters and shearing tools have been developed to cut the reactor internals. All of those equipments are hydraulically driven which is very suitable for submerged applications. The purpose of this paper will be to provide an overview of the Westinghouse mechanical segmentation process, based on actual experience from the work that has been completed to date. (authors)

  12. 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

  13. 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

  14. 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)

  15. 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

  16. 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)

  17. 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.

  18. Design and safety of the Sizewell pressurized water reactor

    International Nuclear Information System (INIS)

    Marshall, W.

    1983-01-01

    The Central Electricity Generating Board propose to build a pressurized water reactor at Sizewell in Suffolk. The PWR Task Force was set up in June 1981 to provide a communications centre for developing firm design proposals for this reactor. These were to follow the Standardized Nuclear Unit Power Plant System designed by Bechtel for the Westinghouse nuclear steam supply system for reactors built in the United States. Changes were required to the design to accommodate, for example, the use of two turbine generators and to satisfy British safety requirements. Differences exist between the British and American licensing procedures. In the UK the statutory responsibility for the safety of a nuclear power station rests unambiguously with the Generating Boards. In the U.S.A. the Nuclear Regulatory Commission issues detailed written instructions, which must be followed precisely. Much of the debate on the safety of nuclear power focuses on the risks of big nuclear accidents. It is necessary to explain to the public what, in a balanced perspective, the risks of accidents actually are. The long-term consequences can be presented in terms of reduction in life expectancy, increased chance of cancer or the equivalent pattern of compulsory cigarette smoking. (author)

  19. Best-estimate methodology for analysis of anticipated transients without scram in pressurized water reactors

    International Nuclear Information System (INIS)

    Rebollo, L.

    1993-01-01

    Union Fenosa, a utility company in Spain, has performed research on pressurized water reactor (PWR) safety with respect to the development of a best-estimate methodology for the analysis of anticipated transients without scram (ATWS), i.e., those anticipated transients for which failure of the reactor protection system is postulated. A scientific and technical approach is adopted with respect to the ATWS phenomenon as it affects a PWR, specifically the Zorita nuclear power plant, a single-loop Westinghouse-designed PWR in Spain. In this respect, an ATWS sequence analysis methodology based on published codes that is generically applicable to any PWR is proposed, which covers all the anticipated phenomena and defines the applicable acceptance criteria. The areas contemplated are cell neutron analysis, core thermal hydraulics, and plant dynamics, which are developed, qualified, and plant dynamics, which are developed, qualified, and validated by comparison with reference calculations and measurements obtained from integral or separate-effects tests

  20. 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.)

  1. 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)

  2. 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.

  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. 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. Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Boiling Water Reactor Benchmark Problem

    Directory of Open Access Journals (Sweden)

    Kulesza Joel A.

    2016-01-01

    Full Text Available This paper describes a comparison of contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a boiling water reactor calculational benchmark problem. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission. The benchmark problem was originally evaluated by Brookhaven National Laboratory using the Oak Ridge National Laboratory discrete ordinates code DORT and the BUGLE-93 cross-section library. In this paper, the Westinghouse RAPTOR-M3G three-dimensional discrete ordinates code was used. A variety of cross-section libraries were used with RAPTOR-M3G including the BUGLE93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory and ALPAN-VII.0 developed at Westinghouse. In comparing the calculated fast reaction rates using the four aforementioned cross-section libraries in the pressure vessel capsule, for six dosimetry reaction rates, a maximum relative difference of 8% was observed. As such, it is concluded that the results calculated by RAPTOR-M3G are consistent with the benchmark and further that the different vintage BUGLE cross-section libraries investigated are largely self-consistent.

  7. 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

  8. 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.

  9. 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

  10. 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.

  11. 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

  12. 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)

  13. 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)

  14. The NCSU [North Carolina State Univ.] freon PWR [pressurized water reactor] loop

    International Nuclear Information System (INIS)

    Caves, J.R.; Doster, J.M.; Miller, G.D.; Wehring, B.W.; Turinsky, P.J.

    1989-01-01

    The nuclear engineering department at North Carolina State University has designed and constructed an operating scale model of a pressurized water reactor (PWR) nuclear steam supply system (NSSS). This facility will be used for education, training, and research. The loop uses electric heaters to simulate the reactor core and Freon as the primary and secondary coolant. Viewing ports at various locations in the loop allow the students to visualize flow regimes in normal and off-normal operating conditions. The objective of the design effort was to scale the thermal-hydraulic characteristics of a two-loop Westinghouse NSSS. Provisions have been made for the simulation of various abnormal occurrences. The model is instrumented in much the same manner as the actual NSSS. Current research projects using the loop include the development of adaptive expert systems to monitor the performance of the facility, diagnose mechanical faults, and to make recommendations to operators for mitigation of accidents. This involves having thermal-hydraulics and core-physics simulators running faster than real time on a mini-supercomputer, with operating parameters updated by communication with the data acquisition and control computer. Further opportunities for research will be investigated as they arise

  15. 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

  16. 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.

  17. 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

  18. High temperature pressure water's blowdown into water. Experimental results

    International Nuclear Information System (INIS)

    Ishida, Toshihisa; Kusunoki, Tsuyoshi; Kasahara, Yoshiyuki; Iida, Hiromasa

    1994-01-01

    The purpose of the present experimental study is to clarify the phenomena in blowdown of high temperature and pressure water in pressure vessel into the containment water for evaluation of design of an advanced marine reactor(MRX). The water blown into the containment water flushed and formed steam jet plume. The steam jet condensed in the water, but some stream penetrated to gas phase of containment and contributed to increase of containment pressure. (author)

  19. A comparison of fuzzy logic-PID control strategies for PWR pressurizer control

    International Nuclear Information System (INIS)

    Kavaklioglu, K.; Ikonomopoulos, A.

    1993-01-01

    This paper describes the results obtained from a comparison performed between classical proportional-integral-derivative (PID) and fuzzy logic (FL) controlling the pressure in a pressurized water reactor (PWR). The two methodologies have been tested under various transient scenarios, and their performances are evaluated with respect to robustness and on-time response to external stimuli. One of the main concerns in the safe operation of PWR is the pressure control in the primary side of the system. In order to maintain the pressure in a PWR at the desired level, the pressurizer component equipped with sprayers, heaters, and safety relief valves is used. The control strategy in a Westinghouse PWR is implemented with a PID controller that initiates either the electric heaters or the sprayers, depending on the direction of the coolant pressure deviation from the setpoint

  20. 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.

  1. 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

  2. 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

  3. 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

  4. Water-Based Pressure-Sensitive Paints

    Science.gov (United States)

    Jordan, Jeffrey D.; Watkins, A. Neal; Oglesby, Donald M.; Ingram, JoAnne L.

    2006-01-01

    Water-based pressure-sensitive paints (PSPs) have been invented as alternatives to conventional organic-solvent-based pressure-sensitive paints, which are used primarily for indicating distributions of air pressure on wind-tunnel models. Typically, PSPs are sprayed onto aerodynamic models after they have been mounted in wind tunnels. When conventional organic-solvent-based PSPs are used, this practice creates a problem of removing toxic fumes from inside the wind tunnels. The use of water-based PSPs eliminates this problem. The waterbased PSPs offer high performance as pressure indicators, plus all the advantages of common water-based paints (low toxicity, low concentrations of volatile organic compounds, and easy cleanup by use of water).

  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. 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

  7. Organic Tank Safety Project: Effect of water partial pressure on the equilibrium water content of waste samples from Hanford Tank 241-U-105

    International Nuclear Information System (INIS)

    Scheele, R.D.; Bredt, P.R.; Sell, R.L.

    1997-09-01

    Water content plays a crucial role in the strategy developed by Webb et al. to prevent propagating or sustainable chemical reactions in the organic-bearing wastes stored in the 20 Organic Tank Watch List tanks at the U.S. Department of Energy''s Hanford Site. Because of water''s importance in ensuring that the organic-bearing wastes continue to be stored safely, Duke Engineering and Services Hanford commissioned the Pacific Northwest National Laboratory to investigate the effect of water partial pressure (P H2O ) on the water content of organic-bearing or representative wastes. Of the various interrelated controlling factors affecting the water content in wastes, P H2O is the most susceptible to being controlled by the and Hanford Site''s environmental conditions and, if necessary, could be managed to maintain the water content at an acceptable level or could be used to adjust the water content back to an acceptable level. Of the various waste types resulting from weapons production and waste-management operations at the Hanford Site, determined that saltcake wastes are the most likely to require active management to maintain the wastes in a Conditionally Safe condition. Webb et al. identified Tank U-105 as a Conditionally Safe saltcake tank. A Conditionally Safe waste is one that is currently safe based on waste classification criteria but could, if dried, be classified as open-quotes Unsafe.close quotes To provide information on the behavior of organic-bearing wastes, the Westinghouse Hanford Company provided us with four waste samples taken from Tank 241-U-105 (U-105) to determine the effect of P H2O on their equilibrium water content

  8. 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.

  9. 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

  10. An estimation of core damage frequency of a pressurized water reactor during midloop operation due to loss of residual heat removal

    International Nuclear Information System (INIS)

    Chao, C.C.; Chen, C.T.; Lee, M.

    1995-01-01

    The core damage frequency caused by loss of residual heat removal (RHR) events was assessed during midloop operation of a Westinghouse-designed three-loop pressurized water reactor. The assessment considers two types of outages (refueling and drained maintenance) and uses failure data collected specifically for shutdown condition. Event trees were developed for five categories of loss of RHR events. Human actions to mitigate the loss of RHR events were identified and human error probabilities were quantified using the human cognitive reliability (HCR) and the technique for human error rate prediction (THERP) models. The results showed that the core damage frequency caused by loss of RHR events during midloop operation was 3.4 x 10 -5 per year. The results also showed that the core damage frequency can be reduced significantly by removing a pressurizer safety valve before entering midloop operation. The establishment of reflux cooling, i.e., decay heat removal through the steam generator secondary side, also plays an important role in mitigating the loss of RHR events during midloop operation

  11. 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.

  12. 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.

  13. 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.

  14. 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

  15. Simulation of low pressure water hammer

    Science.gov (United States)

    Himr, D.; Habán, V.

    2010-08-01

    Numerical solution of water hammer is presented in this paper. The contribution is focused on water hammer in the area of low pressure, which is completely different than high pressure case. Little volume of air and influence of the pipe are assumed in water, which cause sound speed change due to pressure alterations. Computation is compared with experimental measurement.

  16. Chooz A, First Pressurized Water Reactor to be Dismantled in France - 13445

    Energy Technology Data Exchange (ETDEWEB)

    Boucau, Joseph [Westinghouse Electric Company, 43 rue de l' Industrie, Nivelles (Belgium); Mirabella, C. [Westinghouse Electric France, Orsay (France); Nilsson, Lennart [Westinghouse Electric Sweden, Vaesteraas (Sweden); Kreitman, Paul J. [Westinghouse Electric Company, Lake Bluff, IL 60048 (United States); Obert, Estelle [EDF - DPI - CIDEN, Lyon (France)

    2013-07-01

    Nine commercial nuclear power plants have been permanently shut down in France to date, of which the Chooz A plant underwent an extensive decommissioning and dismantling program. Chooz Nuclear Power Station is located in the municipality of Chooz, Ardennes region, in the northeast part of France. Chooz B1 and B2 are 1,500 megawatt electric (MWe) pressurized water reactors (PWRs) currently in operation. Chooz A, a 305 MWe PWR implanted in two caves within a hill, began operations in 1967 and closed in 1991, and will now become the first PWR in France to be fully dismantled. EDF CIDEN (Engineering Center for Dismantling and Environment) has awarded Westinghouse a contract for the dismantling of its Chooz A reactor vessel (RV). The project began in January 2010. Westinghouse is leading the project in a consortium with Nuvia France. The project scope includes overall project management, conditioning of the reactor vessel (RV) head, RV and RV internals segmentation, reactor nozzle cutting for lifting the RV out of the pit and seal it afterwards, dismantling of the RV thermal insulation, ALARA (As Low As Reasonably Achievable) forecast to ensure acceptable doses for the personnel, complementary vacuum cleaner to catch the chips during the segmentation work, needs and facilities, waste characterization and packaging, civil work modifications, licensing documentation. The RV and RV internals will be segmented based on the mechanical cutting technology that Westinghouse applied successfully for more than 13 years. The segmentation activities cover the cutting and packaging plan, tooling design and qualification, personnel training and site implementation. Since Chooz A is located inside two caves, the project will involve waste transportation from the reactor cave through long galleries to the waste buffer area. The project will end after the entire dismantling work is completed, and the waste storage is outside the caves and ready to be shipped either to the ANDRA (French

  17. Flashing of high-pressure saturated water into the pool water

    International Nuclear Information System (INIS)

    Takamasa, Tomoji; Kondo, Koichi; Aya, Izuo.

    1997-01-01

    This paper presents an experimental study on a saturated high-pressure water discharging into a water pool. The purpose of the experiment is to clarify the phenomena that occur by a blow-down of the water from the pressure vessel into the water-filled containment in the case of a wall-crack accident or a LOCA in a passive safety reactor. The results show that a flashing oscillation (FO) occurs when the water discharges into the pool, under specified experimental conditions. The range of the flashing location oscillates between a point very close to and some distance away from the vent hole. The pressures in the vent tube and water pool constantly fluctuate due to the flashing oscillation. The pressure oscillation and alternating flashing location might be caused by the balancing action between the supply of saturated water, flashing at the control volume and steam condensation on the steam-water interface. The frequencies of FO, or frequencies of pressure oscillation and alternating flashing location, increased as water subcooling increased, and as discharging pressure and vent hole diameter decreased. A linear analysis was conducted using a spherical flashing bubble model in which the motion of bubble is controlled by steam condensation. The effects of these parameters on the period of FO in the experiments can be predicted well by the analysis. (author)

  18. 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

  19. Pressure behavior in nuclear reactor containment following a loss of coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Khattab, M; Ibrahim, N A; Bedrose, C D [Reactors department, nuclear research center, atomic energy authority, Cairo, (Egypt)

    1995-10-01

    The scenarios of pressure variation following a loss of coolant accident (LOCA) inside the containment of pressurized water reactor (PWR) have been investigated. Critical mass flow rushing out from high pressure leg through pipe break is used to calculate the rate of coolant. The energy added to the containment atmosphere is determined to specify the rate of growth of pressure and temperature. The seniors of small, medium and large LOCA at 2%, 15%, and 25% flow released are investigated. Safety water spray system is initiated as the pressure reaches the containment design safety limit at about 3 bar to depressurise and to cooldown the system and thereby to reduce the concentration of radioactivity release in the containment atmosphere. The pressure response before and after operation of safety spray system is predicted in each size of LOCA using a typical design of westinghouse PWR system. The heat removal from the containment environment is rejected into the sump by drop-wise condensation mechanism. The effect of initial droplets diameters injected from the nozzles of the spray system is investigated. The results show that the droplet diameter of 3 mm gives best performance. 6 figs.

  20. Pressure behavior in nuclear reactor containment following a loss of coolant accident

    International Nuclear Information System (INIS)

    Khattab, M.; Ibrahim, N.A.; Bedrose, C.D.

    1995-01-01

    The scenarios of pressure variation following a loss of coolant accident (LOCA) inside the containment of pressurized water reactor (PWR) have been investigated. Critical mass flow rushing out from high pressure leg through pipe break is used to calculate the rate of coolant. The energy added to the containment atmosphere is determined to specify the rate of growth of pressure and temperature. The seniors of small, medium and large LOCA at 2%, 15%, and 25% flow released are investigated. Safety water spray system is initiated as the pressure reaches the containment design safety limit at about 3 bar to depressurise and to cooldown the system and thereby to reduce the concentration of radioactivity release in the containment atmosphere. The pressure response before and after operation of safety spray system is predicted in each size of LOCA using a typical design of westinghouse PWR system. The heat removal from the containment environment is rejected into the sump by drop-wise condensation mechanism. The effect of initial droplets diameters injected from the nozzles of the spray system is investigated. The results show that the droplet diameter of 3 mm gives best performance. 6 figs

  1. 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

  2. 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

  3. Pressurized water reactor fuel performance problems connected with fuel cladding corrosion processes

    International Nuclear Information System (INIS)

    Dobrevski, I.; Zaharieva, N.

    2008-01-01

    Generally, Pressurized Water Reactor (WWER, PWR) Fuel Element Performance is connected with fuel cladding corrosion and crud deposition processes. By transient to extended fuel cycles in nuclear power reactors, aiming to achieve higher burnup and better fuel utilization, the role of these processes increases significantly. This evolution modifies the chemical and electrochemical conditions in the reactor primary system, including change of fuel claddings' environment. The higher duty cores are always attended with increased boiling (sub-cooled nucleate boiling) mainly on the feed fuel assemblies. This boiling process on fuel cladding surfaces can cause different consequences on fuel element cladding's environment characteristics. In the case of boiling at the cladding surfaces without or with some cover of corrosion product deposition, the behavior of gases dissolved in water phase is strongly influenced by the vapor generation. The increase of vapor partial pressure will reduce the partial pressures of dissolved gases and will cause their stripping out. By these circumstances the concentrations of dissolved gases in cladding wall water layer can dramatically decrease, including also the case by which all dissolved gases to be stripped out. On the other hand it is known that the hydrogen is added to primary coolant in order to avoid the production of oxidants by radiolysis of water. It is clear that if boiling strips out dissolved hydrogen, the creation of oxidizing conditions at the cladding surfaces will be favored. In this case the local production of oxidants will be a result from local processes of water radiolysis, by which not only both oxygen (O 2 ) and hydrogen (H 2 ) but also hydrogen peroxide (H 2 O 2 ) will be produced. While these hydrogen and oxygen will be stripped out preferentially by boiling, the bigger part of hydrogen peroxide will remain in wall water phase and will act as the most important factor for creation of oxidizing conditions in fuel

  4. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-01

    The use of internally and externally cooled annular fuel rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and economic assessment. The investigation was conducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperature. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasibility issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density

  5. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31

    The use of internally and externally cooled annular fule rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and econmic assessment. The investigation was donducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperatre. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasiblity issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density.

  6. Emergency operating procedures guidelines for pressurized water reactors - a progress report

    International Nuclear Information System (INIS)

    Lyon, W.C.

    1984-01-01

    Emergency Operating Procedures (EOPs) contain the instructions the operator will follow to control a nuclear plant whenever a condition exists that potentially jeopardizes the fuel cladding, the reactor coolant system (RCS) pressure boundary, or the containment. The EOPs are prepared from guidelines which contain the major operator instructions that will be in the EOPs. Guidelines have been prepared by owners' groups having Babcock and Wilcox (BandW), Combustion Engineering (CE), General Electric (GE), and Westinghouse (W) plants. These guidelines cover many aspects of full power operation. Future effort is anticipated to complete coverage of transient events, including severe accidents, all power conditions, and shutdown. This paper describes the philosophy which has guided NRC technical review of guidelines, progress achieved in providing comprehensive coverage of emergency conditions for PWRs, and anticipated future technical activities

  7. Pressurized-water reactors

    International Nuclear Information System (INIS)

    Bush, S.H.

    1983-03-01

    An overview of the pressurized-water reactor (PWR) pressure boundary problems is presented. Specifically exempted will be discussions of problems with pumps, valves and steam generators on the basis that they will be covered in other papers. Pressure boundary reliability is examined in the context of real or perceived problems occurring over the past 5 to 6 years since the last IAEA Reliability Symposium. Issues explicitly covered will include the status of the pressurized thermal-shock problem, reliability of inservice inspections with emphasis on examination of the region immediately under the reactor pressure vessel (RPV) cladding, history of piping failures with emphasis on failure modes and mechanisms. Since nondestructive examination is the topic of one session, discussion will be limited to results rather than techniques

  8. Detailed pressure drop measurements in single-and two-phase adiabatic air-water turbulent flows in realistic BWR fuel assembly geometry with spacer grids

    International Nuclear Information System (INIS)

    Caraghiaur, Diana; Frid, Wiktor; Tillmark, Nils

    2004-01-01

    In recent years, advance numerical simulation tools based on CFD methods have been increasingly used in various multi-phase flow applications. One of these is two-phase flow in fuel assemblies of Boiling Water Reactors. The important and often missing aspect of this development is validation of CFD codes against proper experimental data. The purpose of the current paper is to present detailed pressure measurements over a spacer grid in low pressure adiabatic single- and bubbly two-phase flow, which will be used to further develop a CFD code for BWR fuel bundle analysis. The experiments have been carried out in a n asymmetric 24-rod sub-bundle, representing one quarter of a Westinghouse SVEA-96 nuclear reactor fuel assembly. Single-phase flow measurements have been performed at superficial velocities between 0.90-4.50 m/s and in the two-phase flow, which was simulated by air-water mixture, measurements have been performed at void fractions ranging from 4 to 12% and liquid superficial velocity of 4.50 m/s. In order to increase the number of measuring points, five pressure taps were drilled in one of the rods, which was easily moved vertically by a traversing system, covering most of the points in axial direction. Any of the rods in the bundle could be substitute by the pressure sensing rod and the measurements were made for five pressure taps facing-angles. A detailed pressure distribution comparison between single- and two-phase flows for different sub-channel positions and different flow conditions was performed over one of the spacers. In addition, single-phase pressure drop measurements in the upper part of the test section comprising two spacer grids have been carried out. (author)

  9. High pressure experimental water loop

    International Nuclear Information System (INIS)

    Grenon, M.

    1958-01-01

    A high pressure experimental water loop has been made for studying the detection and evolution of cladding failure in a pressurized reactor. The loop has been designed for a maximum temperature of 360 deg. C, a maximum of 160 kg/cm 2 and flow rates up to 5 m 3 /h. The entire loop consists of several parts: a main circuit with a canned rotor circulation pump, steam pressurizer, heating tubes, two hydro-cyclones (one de-gasser and one decanter) and one tubular heat exchanger; a continuous purification loop, connected in parallel, comprising pressure reducing valves and resin pots which also allow studies of the stability of resins under pressure, temperature and radiation; following the gas separator is a gas loop for studying the recombination of the radiolytic gases in the steam phase. The preceding circuits, as well as others, return to a low pressure storage circuit. The cold water of the low pressure storage flask is continuously reintroduced into the high pressure main circuit by means of a return pump at a maximum head of 160 kg /cm 2 , and adjusted to the pressurizer level. This loop is also a testing bench for the tight high pressure apparatus. The circulating pump and the connecting flanges (Oak Ridge type) are water-tight. The feed pump and the pressure reducing valves are not; the un-tight ones have a system of leak recovery. To permanently check the tightness the circuit has been fitted with a leak detection system (similar to the HRT one). (author) [fr

  10. 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)

  11. 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)

  12. 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

  13. 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.

  14. 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.

  15. 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

  16. New lineup of light water reactors

    International Nuclear Information System (INIS)

    Okamura, Kiyoshi; Oshima, Koichiro; Kitsukawa, Keisuke

    2007-01-01

    Toshiba is promoting technical studies for upcoming nuclear power plants based on its large accumulation of experience in boiling water reactor (BWR) design, manufacturing, construction, and maintenance. Our goal is to achieve higher reliability, lower life-cycle costs, and better competitiveness for nuclear power plants compared with other energy sources. In addition, we are developing a new light water reactor (LWR) lineup featuring the safest and most economical LWRs in the world as next-generation reactors almost at new construction and replacement in the Japanese and international markets expected to start from the 2020s. We are committed not only to developing BWRs with the world's highest performance but also to participating in the pressurized water reactor (PWR) market, taking advantage of the synergistic effect of both Toshiba's and Westinghouse's experience. (author)

  17. 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

  18. Water-Pressure Distribution on Seaplane Float

    Science.gov (United States)

    Thompson, F L

    1929-01-01

    The investigation presented in this report was conducted for the purpose of determining the distribution and magnitude of water pressures likely to be experienced on seaplane hulls in service. It consisted of the development and construction of apparatus for recording water pressures lasting one one-hundredth second or longer and of flight tests to determine the water pressures on a UO-1 seaplane float under various conditions of taxiing, taking off, and landing. The apparatus developed was found to operate with satisfactory accuracy and is suitable for flight tests on other seaplanes. The tests on the UO-1 showed that maximum pressures of about 6.5 pounds per square inch occur at the step for the full width of the float bottom. Proceeding forward from the step the maximum pressures decrease in magnitude uniformly toward the bow, and the region of highest pressures narrows toward the keel. Immediately abaft the step the maximum pressures are very small, but increase in magnitude toward the stern and there once reached a value of about 5 pounds per square inch. (author)

  19. 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

  20. High Pressure Industrial Water Facility

    Science.gov (United States)

    1992-01-01

    In conjunction with Space Shuttle Main Engine testing at Stennis, the Nordberg Water Pumps at the High Pressure Industrial Water Facility provide water for cooling the flame deflectors at the test stands during test firings.

  1. Contribution of water vapor pressure to pressurization of plutonium dioxide storage containers

    Science.gov (United States)

    Veirs, D. Kirk; Morris, John S.; Spearing, Dane R.

    2000-07-01

    Pressurization of long-term storage containers filled with materials meeting the US DOE storage standard is of concern.1,2 For example, temperatures within storage containers packaged according to the standard and contained in 9975 shipping packages that are stored in full view of the sun can reach internal temperatures of 250 °C.3 Twenty five grams of water (0.5 wt.%) at 250 °C in the storage container with no other material present would result in a pressure of 412 psia, which is limited by the amount of water. The pressure due to the water can be substantially reduced due to interactions with the stored material. Studies of the adsorption of water by PuO2 and surface interactions of water with PuO2 show that adsorption of 0.5 wt.% of water is feasible under many conditions and probable under high humidity conditions.4,5,6 However, no data are available on the vapor pressure of water over plutonium dioxide containing materials that have been exposed to water.

  2. PWR [pressurized water reactor] pressurizer transient response: Final report

    International Nuclear Information System (INIS)

    Murphy, S.I.

    1987-08-01

    To predict PWR pressurizer transients, Ahl proposed a three region model with a universal coefficient to represent condensation on the water surface. Specifically, this work checks the need for three regions and the modeling of the interfacial condensation coefficient. A computer model has been formulated using the basic mass and energy conservation laws. A two region vapor and liquid model was first used to predict transients run on a one-eleventh scale Freon pressurizer. These predictions verified the need for a second liquid region. As a result, a three region model was developed and used to predict full-scale pressurizer transients at TMI-2, Shippingport, and Stade. Full-scale pressurizer predictions verified the three region model and pointed out the shortcomings of Ahl's universal condensation coefficient. In addition, experiments were run using water at low pressure to study interface condensation. These experiments showed interface condensation to be significant only when spray flow is turned on; this result was incorporated in the final three region model

  3. 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

  4. 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.)

  5. 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.

  6. A Mixed-Oxide Assembly Design for Rapid Disposition of Weapons Plutonium in Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Alonso, Gustavo; Adams, Marvin L.

    2002-01-01

    We have created a new mixed-oxide (MOX) fuel assembly design for standard pressurized water reactors (PWRs). Design goals were to maximize the plutonium throughput while introducing the lowest perturbation possible to the control and safety systems of the reactor. Our assembly design, which we call MIX-33, offers some advantages for the disposition of weapons-grade plutonium; it increases the disposition rate by 8% while increasing the worth of control material, compared to a previous Westinghouse design. The MIX-33 design is based upon two ideas: the use of both uranium and plutonium fuel pins in the same assembly, and the addition of water holes in the assembly. The main result of this paper is that both of these ideas are effective at increasing Pu throughput and increasing the worth of control material. With this new design, according to our analyses, we can transition smoothly from a full low-enriched-uranium (LEU) core to a full MIX-33 core while meeting the operational and safety requirements of a standard PWR. Given an interruption of the MOX supply, we can transition smoothly back to full LEU while meeting safety margins and using standard LEU assemblies with uniform pinwise enrichment distribution. If the MOX supply is interrupted for only one cycle, the transition back to a full MIX-33 core is not as smooth; high peaking could cause power to be derated by a few percent for a few weeks at the beginning of one transition cycle

  7. Water-Based Pressure Sensitive Paint

    Science.gov (United States)

    Oglesby, Donald M.; Ingram, JoAnne L.; Jordan, Jeffrey D.; Watkins, A. Neal; Leighty, Bradley D.

    2004-01-01

    Preparation and performance of a water-based pressure sensitive paint (PSP) is described. A water emulsion of an oxygen permeable polymer and a platinum porphyrin type luminescent compound were dispersed in a water matrix to produce a PSP that performs well without the use of volatile, toxic solvents. The primary advantages of this PSP are reduced contamination of wind tunnels in which it is used, lower health risk to its users, and easier cleanup and disposal. This also represents a cost reduction by eliminating the need for elaborate ventilation and user protection during application. The water-based PSP described has all the characteristics associated with water-based paints (low toxicity, very low volatile organic chemicals, and easy water cleanup) but also has high performance as a global pressure sensor for PSP measurements in wind tunnels. The use of a water-based PSP virtually eliminates the toxic fumes associated with the application of PSPs to a model in wind tunnels.

  8. Westinghouse AP1000 Electrical Generation Costs - Meeting Marketplace Requirements

    International Nuclear Information System (INIS)

    Paulson, C. Keith

    2002-01-01

    The re-emergence of nuclear power as a leading contender for new base-load electrical generation is not an occurrence of happenstance. The nuclear industry, in general, and Westinghouse, specifically, have worked diligently with the U.S. power companies and other nuclear industry participants around the world to develop future plant designs and project implementation models that address prior problem areas that led to reduced support for nuclear power. In no particular order, the issues that Westinghouse, as an engineering and equipment supply company, focused on were: safety, plant capital costs, construction schedule reductions, plant availability, and electric generation costs. An examination of the above criteria quickly led to the conclusion that as long as safety is not compromised, simplifying plant designs can lead to positive progress of the desired endpoints for the next and later generations of nuclear units. The distinction between next and later generations relates to the readiness of the plant design for construction implementation. In setting requirement priorities, one axiom is inviolate: There is no exception, nor will there be, to the Golden Rule of business. In the electric power generation industry, once safety goals are met, low generation cost is the requirement that rules, without exception. The emphasis in this paper on distinguishing between next and later generation reactors is based on the recognition that many designs have been purposed for future application, but few have been able to attain the design pedigree required to successfully meet the requirements for next generation nuclear units. One fact is evident: Another generation of noncompetitive nuclear plants will cripple the potential for nuclear to take its place as a major contributor to new electrical generation. Only two plant designs effectively meet the economic tests and demonstrate both unparalleled safety and design credibility due to extensive progress toward engineering

  9. 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.

  10. High-pressure water facility

    Science.gov (United States)

    2006-01-01

    NASA Test Operations Group employees, from left, Todd Pearson, Tim Delcuze and Rodney Wilkinson maintain a water pump in Stennis Space Center's high-pressure water facility. The three were part of a group of employees who rode out Hurricane Katrina at the facility and helped protect NASA's rocket engine test complex.

  11. 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

  12. 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

  13. Pressure pressure-balanced pH sensing system for high temperature and high pressure water

    International Nuclear Information System (INIS)

    Tachibana, Koji

    1995-01-01

    As for the pH measurement system for high temperature, high pressure water, there have been the circumstances that first the reference electrodes for monitoring corrosion potential were developed, and subsequently, it was developed for the purpose of maintaining the soundness of metallic materials in high temperature, high pressure water in nuclear power generation. In the process of developing the reference electrodes for high temperature water, it was clarified that the occurrence of stress corrosion cracking in BWRs is closely related to the corrosion potential determined by dissolved oxygen concentration. As the types of pH electrodes, there are metal-hydrogen electrodes, glass electrodes, ZrO 2 diaphragm electrodes and TiO 2 semiconductor electrodes. The principle of pH measurement using ZrO 2 diaphragms is explained. The pH measuring system is composed of YSZ element, pressure-balanced type external reference electrode, pressure balancer and compressed air vessel. The stability and pH response of YSZ elements are reported. (K.I.)

  14. Managing water pressure for water savings in developing countries

    African Journals Online (AJOL)

    2014-03-03

    Mar 3, 2014 ... effort into providing customers with a reliable level of service, often via poor water ... budgets. There are many factors contributing to water losses in water .... given relationship does not reflect the impact of pressure on.

  15. 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

  16. Pressurized water-reactor feedwater piping response to water hammer

    International Nuclear Information System (INIS)

    Arthur, D.

    1978-03-01

    The nuclear power industry is interested in steam-generator water hammer because it has damaged the piping and components at pressurized water reactors (PWRs). Water hammer arises when rapid steam condensation in the steam-generator feedwater inlet of a PWR causes depressurization, water-slug acceleration, and slug impact at the nearest pipe elbow. The resulting pressure pulse causes the pipe system to shake, sometimes violently. The objective of this study is to evaluate the potential structural effects of steam-generator water hammer on feedwater piping. This was accomplished by finite-element computation of the response of two sections of a typical feedwater pipe system to four representative water-hammer pulses. All four pulses produced high shear and bending stresses in both sections of pipe. Maximum calculated pipe stresses varied because the sections had different characteristics and were sensitive to boundary-condition modeling

  17. 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

  18. 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)

  19. 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

  20. 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.

  1. 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

  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. 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%

  5. 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)

  6. Water cooled static pressure probe

    Science.gov (United States)

    Lagen, Nicholas T. (Inventor); Eves, John W. (Inventor); Reece, Garland D. (Inventor); Geissinger, Steve L. (Inventor)

    1991-01-01

    An improved static pressure probe containing a water cooling mechanism is disclosed. This probe has a hollow interior containing a central coolant tube and multiple individual pressure measurement tubes connected to holes placed on the exterior. Coolant from the central tube symmetrically immerses the interior of the probe, allowing it to sustain high temperature (in the region of 2500 F) supersonic jet flow indefinitely, while still recording accurate pressure data. The coolant exits the probe body by way of a reservoir attached to the aft of the probe. The pressure measurement tubes are joined to a single, larger manifold in the reservoir. This manifold is attached to a pressure transducer that records the average static pressure.

  7. Flaw evaluation of pressure vessel in pressurized water reactor

    International Nuclear Information System (INIS)

    Park, Ki Sung; Kim, Min Geol; Jeon, Chae Hong; Rhim, Soon Hyung; Kim, Seung Tae

    1999-01-01

    Flaw evaluation should be performed to determine the acceptance of a surface or a subsurface flaw detected during the in-service inspection without any repair or replacement. In this paper, the evaluation methodology and procedure were established according to ASME code Sec. XI and the evaluation program was coded. Using this program, a field engineer who doesn't have enough knowledge on fracture mechanics may be able to perform prompt and accurate flaw evaluation on site and decide whether a detected flaw be allowable or not. Analysis results were compared with those obtained from Westinghouse program called KCAL and FCG. Both results made good agreement and accuracy of the program developed in this paper was verified.=20

  8. Final report on 3-D experiment project air-water upper plenum experiments

    International Nuclear Information System (INIS)

    Jacoby, J.K.; Mohr, C.M.

    1978-11-01

    The results are presented from upper plenum air-water reflood behavior testing performed as part of the program to investigate three-dimensional aspects of PWR LOCA research. Tests described were performed at near ambient temperature and pressure in a plexiglass vessel which included the important features of the upper core and upper plenum regions corresponding to a single fuel bundle in both Westinghouse Electric Corporation (Trojan) and Kraftwerk Union (KKU) PWR designs. The data included observed two-phase flow characteristics, particularly with regard to countercurrent flow, and cinematography of the characteristic upper plenum flow patterns

  9. Water Pressure Distribution on a Flying Boat Hull

    Science.gov (United States)

    Thompson, F L

    1931-01-01

    This is the third in a series of investigations of the water pressures on seaplane floats and hulls, and completes the present program. It consisted of determining the water pressures and accelerations on a Curtiss H-16 flying boat during landing and taxiing maneuvers in smooth and rough water.

  10. Fuzzy control applied to nuclear power plant pressurizer system

    International Nuclear Information System (INIS)

    Oliveira, Mauro V.; Almeida, Jose C.S.

    2011-01-01

    In a pressurized water reactor (PWR) nuclear power plants (NPPs) the pressure control in the primary loop is very important for keeping the reactor in a safety condition and improve the generation process efficiency. The main component responsible for this task is the pressurizer. The pressurizer pressure control system (PPCS) utilizes heaters and spray valves to maintain the pressure within an operating band during steady state conditions, and limits the pressure changes, during transient conditions. Relief and safety valves provide overpressure protection for the reactor coolant system (RCS) to ensure system integrity. Various protective reactor trips are generated if the system parameters exceed safe bounds. Historically, a proportional-integral derivative (PID) controller is used in PWRs to keep the pressure in the set point, during those operation conditions. The purpose of this study has two main goals: first is to develop a pressurizer model based on artificial neural networks (ANNs); second is to develop a fuzzy controller for the PWR pressurizer pressure, and compare its performance with the P controller. Data from a simulator PWR plant was used to test the ANN and the controllers as well. The reference simulator is a Westinghouse 3-loop PWR plant with a total thermal output of 2785 MWth. The simulation results show that the pressurizer ANN model response are in reasonable agreement with the simulated power plant, and the fuzzy controller built in this study has better performance compared to the P controller. (author)

  11. Fuzzy control applied to nuclear power plant pressurizer system

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Mauro V.; Almeida, Jose C.S., E-mail: mvitor@ien.gov.b, E-mail: jcsa@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    In a pressurized water reactor (PWR) nuclear power plants (NPPs) the pressure control in the primary loop is very important for keeping the reactor in a safety condition and improve the generation process efficiency. The main component responsible for this task is the pressurizer. The pressurizer pressure control system (PPCS) utilizes heaters and spray valves to maintain the pressure within an operating band during steady state conditions, and limits the pressure changes, during transient conditions. Relief and safety valves provide overpressure protection for the reactor coolant system (RCS) to ensure system integrity. Various protective reactor trips are generated if the system parameters exceed safe bounds. Historically, a proportional-integral derivative (PID) controller is used in PWRs to keep the pressure in the set point, during those operation conditions. The purpose of this study has two main goals: first is to develop a pressurizer model based on artificial neural networks (ANNs); second is to develop a fuzzy controller for the PWR pressurizer pressure, and compare its performance with the P controller. Data from a simulator PWR plant was used to test the ANN and the controllers as well. The reference simulator is a Westinghouse 3-loop PWR plant with a total thermal output of 2785 MWth. The simulation results show that the pressurizer ANN model response are in reasonable agreement with the simulated power plant, and the fuzzy controller built in this study has better performance compared to the P controller. (author)

  12. Where Did the Water Go?: Boyle's Law and Pressurized Diaphragm Water Tanks

    Science.gov (United States)

    Brimhall, James; Naga, Sundar

    2007-01-01

    Many homes use pressurized diaphragm tanks for storage of water pumped from an underground well. These tanks are very carefully constructed to have separate internal chambers for the storage of water and for the air that provides the pressure. One might expect that the amount of water available for use from, for example, a 50-gallon tank would be…

  13. Westinghouse Hanford Company package testing capabilities

    International Nuclear Information System (INIS)

    Hummer, J.H.; Mercado, M.S.

    1993-07-01

    The Department of Energy's Hanford Site is a 1,450-km 2 (560-mi 2 ) installation located in southeastern Washington State. Established in 1943 as a plutonium production facility, Hanford's role has evolved into one of environmental restoration and remediation. Many of these environmental restoration and remediation activities involve transportation of radioactive/hazardous materials. Packagings used for the transportation of radioactive/hazardous materials must be capable of meeting certain normal transport and hypothetical accident performance criteria. Evaluations of performance to these criteria typically involve a combination of analysis and testing. Required tests may include the free drop, puncture, penetration, compression, thermal, heat, cold, vibration, water spray, water immersion, reduced pressure, and increased pressure tests. The purpose of this paper is to outline the Hanford capabilities for performing each of these tests

  14. 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.

  15. 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

  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. 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.

  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. Pressurized water reactor flow arrangement

    International Nuclear Information System (INIS)

    Gibbons, J.F.; Knapp, R.W.

    1980-01-01

    A flow path is provided for cooling the control rods of a pressurized water reactor. According to this scheme, a small amount of cooling water enters the control rod guide tubes from the top and passes downwards through the tubes before rejoining the main coolant flow and passing through the reactor core. (LL)

  20. Solubility and physical properties of sugars in pressurized water

    International Nuclear Information System (INIS)

    Saldaña, Marleny D.A.; Alvarez, Víctor H.; Haldar, Anupam

    2012-01-01

    Highlights: ► Sugar solubility in pressurized water and density at high pressures were measured. ► Glucose solubility was higher than that of lactose as predicted by their σ-profiles. ► Sugar aqueous solubility decreased with an increase in pressure from 15 to 120 bar. ► Aqueous glucose molecular packing shows high sensitivity to pressure. ► The COSMO-SAC model qualitatively predicted the sugar solubility data. - Abstract: In this study, the solubility, density, and refractive index of glucose and lactose in water as a function of temperature were measured. For solubility of sugars in pressurized water, experimental data were obtained at pressures of (15 to 120) bar and temperatures of (373 to 433) K using a dynamic flow high pressure system. Density data for aqueous sugar solutions were obtained at pressures of (1 to 300) bar and temperatures of (298 to 343) K. The refractive index of aqueous sugar solutions was obtained at 293 K and atmospheric pressure. Activity coefficient models, Van Laar and the Conductor-like Screening Model-Segment Activity Coefficient (COSMO-SAC), were used to fit and predict the experimental solubility data, respectively. The results obtained showed that the solubility of both sugars in pressurized water increase with an increase in temperature. However, with the increase of pressure from 15 bar to 120 bar, the solubility of both sugars in pressurized water decreased. The Van Laar model fit the experimental aqueous solubility data with deviations lower than 13 and 53% for glucose and lactose, respectively. The COSMO-SAC model predicted qualitatively the aqueous solubility of these sugars.

  1. 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

  2. International feedback experience on the cutting of reactor internal components

    International Nuclear Information System (INIS)

    Boucau, J.

    2014-01-01

    Westinghouse capitalizes more than 30 years of experience in the cutting of internal components of reactor and their packaging in view of their storage. Westinghouse has developed and validated different methods for cutting: plasma torch cutting, high pressure abrasive water jet cutting, electric discharge cutting and mechanical cutting. A long feedback experience has enabled Westinghouse to list the pros and cons of each cutting technology. The plasma torch cutting is fast but rises dosimetry concerns linked to the control of the cuttings and the clarity of water. Abrasive water jet cutting requires the installation of costly safety devices and of an equipment for filtering water but this technology allows accurate cuttings in hard-to-reach zones. Mechanical cutting is the most favourable technology in terms of wastes generation and of the clarity of water but the cutting speed is low. (A.C.)

  3. High-pressure-induced water penetration into 3-isopropylmalate dehydrogenase

    International Nuclear Information System (INIS)

    Nagae, Takayuki; Kawamura, Takashi; Chavas, Leonard M. G.; Niwa, Ken; Hasegawa, Masashi; Kato, Chiaki; Watanabe, Nobuhisa

    2012-01-01

    Structures of 3-isopropylmalate dehydrogenase were determined at pressures ranging from 0.1 to 650 MPa. Comparison of these structures gives a detailed picture of the swelling of a cavity at the dimer interface and the generation of a new cleft on the molecular surface, which are accompanied by water penetration. Hydrostatic pressure induces structural changes in proteins, including denaturation, the mechanism of which has been attributed to water penetration into the protein interior. In this study, structures of 3-isopropylmalate dehydrogenase (IPMDH) from Shewanella oneidensis MR-1 were determined at about 2 Å resolution under pressures ranging from 0.1 to 650 MPa using a diamond anvil cell (DAC). Although most of the protein cavities are monotonically compressed as the pressure increases, the volume of one particular cavity at the dimer interface increases at pressures over 340 MPa. In parallel with this volume increase, water penetration into the cavity could be observed at pressures over 410 MPa. In addition, the generation of a new cleft on the molecular surface accompanied by water penetration could also be observed at pressures over 580 MPa. These water-penetration phenomena are considered to be initial steps in the pressure-denaturation process of IPMDH

  4. Adaptive Reference Control for Pressure Management in Water Networks

    DEFF Research Database (Denmark)

    Kallesøe, Carsten; Jensen, Tom Nørgaard; Wisniewski, Rafal

    2015-01-01

    Water scarcity is an increasing problem worldwide and at the same time a huge amount of water is lost through leakages in the distribution network. It is well known that improved pressure control can lower the leakage problems. In this work water networks with a single pressure actuator and several....... Subsequently, these relations are exploited in an adaptive reference control scheme for the actuator pressure that ensures constant pressure at the critical points. Numerical experiments underpin the results. © Copyright IEEE - All rights reserved....

  5. Assessment of water pipes durability under pressure surge

    Science.gov (United States)

    Pham Ha, Hai; Minh, Lanh Pham Thi; Tang Van, Lam; Bulgakov, Boris; Bazhenova, Soafia

    2017-10-01

    Surge phenomenon occurs on the pipeline by the closing valve or pump suddenly lost power. Due to the complexity of the water hammer simulation, previous researches have only considered water hammer on the single pipe or calculation of some positions on water pipe network, it have not been analysis for all of pipe on the water distribution systems. Simulation of water hammer due to closing valve on water distribution system and the influence level of pressure surge is evaluated at the defects on pipe. Water hammer on water supply pipe network are simulated by Water HAMMER software academic version and the capacity of defects are calculated by SINTAP. SINTAP developed from Brite-Euram projects in Brussels-Belgium with the aim to develop a process for assessing the integrity of the structure for the European industry. Based on the principle of mechanical fault, indicating the size of defects in materials affect the load capacity of the product in the course of work, the process has proposed setting up the diagram to fatigue assessment defect (FAD). The methods are applied for water pipe networks of Lien Chieu district, Da Nang city, Viet Nam, the results show the affected area of wave pressure by closing the valve and thereby assess the greatest pressure surge effect to corroded pipe. The SINTAP standard and finite element mesh analysis at the defect during the occurrence of pressure surge which will accurately assess the bearing capacity of the old pipes. This is one of the bases to predict the leakage locations on the water distribution systems. Amount of water hammer when identified on the water supply networks are decreasing due to local losses at the nodes as well as the friction with pipe wall, so this paper adequately simulate water hammer phenomena applying for actual water distribution systems. The research verified that pipe wall with defect is damaged under the pressure surge value.

  6. Swelling pressure and water absorption property of compacted granular bentonite during water absorption

    International Nuclear Information System (INIS)

    Oyamada, T.; Komine, H.; Murakami, S.; Sekiguchi, T.; Sekine, I.

    2012-01-01

    Document available in extended abstract form only. Bentonite is currently planned to be used as buffer materials in engineered barrier of radioactive waste disposal. Granular bentonites are expected as the materials used in constructions as buffer materials by in-situ compaction methods. After applying these buffer materials, it is expected that the condition of the buffer area changes in long-term by the seepage of groundwater into buffer area. Therefore, it is important to understand water movement and swelling behavior of the buffer materials for evaluating the performance of engineered barrier. In this study, we investigated water absorption property and swelling pressure of compacted granular bentonite. Specifically, the process of swelling pressure and amount of water absorption of granular bentonite-GX (Kunigel-GX, produced at the Tsukinuno mine in Japan) were observed by laboratory tests. To discuss the influence of maximum grain size of bentonite particle on swelling pressure and water absorption property, two types of samples were used. One is granular sample which is Bentonite-GX controlled under 2 mm the maximum grain size, the other is milled sample which is Bentonite-GX with the maximum grain size under 0.18 mm by milling with the agate mortar. In addition, the mechanism on the swelling pressure of compacted granular bentonite was considered and discussed. In the cases of granular sample, swelling pressure increases rapidly, then gradually continues to increase up to maximum value. In the cases of milled sample, swelling pressure also increases rapidly at first. However, then its value decreases before progressing of gradual increase continues. Especially, this trend was clearly observed at a relatively low dry density. At the peaks of these curves, the swelling pressure of granular samples is lower than that of milled samples. In addition, the increasing of swelling pressure by the time the peak observed during the process of swelling pressure from

  7. Water Pressure Distribution on a Twin-Float Seaplane

    Science.gov (United States)

    Thompson, F L

    1930-01-01

    This is the second of a series of investigations to determine water pressure distribution on various types of seaplane floats and hulls, and was conducted on a twin-float seaplane. It consisted of measuring water pressures and accelerations on a TS-1 seaplane during numerous landing and taxiing maneuvers at various speeds and angles. The results show that water pressures as great as 10 lbs. per sq. in.may occur at the step in various maneuvers and that pressures of approximately the same magnitude occur at the stern and near the bow in hard pancake landings with the stern way down. At the other parts of the float the pressures are less and are usually zero or slightly negative for some distance abaft the step. A maximum negative pressure of 0.87 lb. Per square inch was measured immediately abaft the step. The maximum positive pressures have a duration of approximately one-twentieth to one-hundredth second at any given location and are distributed over a very limited area at any particular instant.

  8. Pressurized-water coolant nuclear reactor steam generator

    International Nuclear Information System (INIS)

    Mayer, H.; Schroder, H.J.

    1975-01-01

    A description is given of a pressurized-water coolant nuclear reactor steam generator having a vertical housing for the steam generating water and containing an upstanding heat exchanger to which the pressurized-water coolant passes and which is radially surrounded by a guide jacket supporting a water separator on its top. By thermosiphon action the steam generating water flows upward through and around the heat exchanger within the guide chamber to the latter's top from which it flows radially outwardly and downwardly through a down draft space formed between the outside of the jacket and the housing. The water separator discharges separated water downwardly. The housing has a feedwater inlet opening adjacent to the lower portion of the heat exchanger, providing preheating of the introduced feedwater. This preheated feedwater is conveyed by a duct upwardly to a location where it mixes with the water discharged from the water separator

  9. 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

  10. 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

  11. 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

  12. Containment pressure analysis model using CONTEMPT-LT

    International Nuclear Information System (INIS)

    Gupta, R.N.

    1975-09-01

    An analytical model for evaluating the reactor containment pressure transient following a loss-of-coolant accident (LOCA) is presented. The model uses the CONTEMPT-LT computer program developed by Aerojet Nuclear Company. The sample problem studied is the containment response following the most severe postulated LOCA at the Yankee Rowe Nuclear Power Station. The results show good agreement with the response predicted by Westinghouse Electric Corporation. (auth)

  13. Low cost sonoluminescence experiment in pressurized water

    Energy Technology Data Exchange (ETDEWEB)

    Bernal, L; Insabella, M [LADOP, University of Mar del Plata (Argentina); Bilbao, L [INFIP, University of Buenos Aires and CONICET (Argentina)

    2012-06-19

    We present a low cost design for demostration and mesurements of light emission from a sonoluminescence experiment. Using pressurized water introduced in an acrylic cylinder and one piezoelectric from an ultrasonic cleaner, we are able to generate cavitacion zones with emission of light. The use of argon to pressurize the water improves the emission an the light can be seen at naked eye in a softlit ambient.

  14. Low cost sonoluminescence experiment in pressurized water

    International Nuclear Information System (INIS)

    Bernal, L; Insabella, M; Bilbao, L

    2012-01-01

    We present a low cost design for demostration and mesurements of light emission from a sonoluminescence experiment. Using pressurized water introduced in an acrylic cylinder and one piezoelectric from an ultrasonic cleaner, we are able to generate cavitacion zones with emission of light. The use of argon to pressurize the water improves the emission an the light can be seen at naked eye in a softlit ambient.

  15. 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.

  16. Heat insulation device for reactor pressure vessel in water

    International Nuclear Information System (INIS)

    Nakamura, Heiichiro; Tanaka, Yoshimi.

    1993-01-01

    Outer walls of a reactor pressure vessel are covered with water-tight walls made of metals. A heat insulation metal material is disposed between them. The water tight walls are joined by welding and flanges. A supply pipeline for filling gases and a discharge pipeline are in communication with the inside of the water tight walls. Further, a water detector is disposed in the midway of the gas discharge pipeline. With such a constitution, the following advantages can be attained. (1) Heat transfer from the reactor pressure vessel to water of a reactor container can be suppressed by filled gases and heat insulation metal material. (2) Since the pressure at the inside of the water tight walls can be equalized with the pressure of the inside of the reactor container, the thickness of the water-tight walls can be reduced. (3) Since intrusion of water to the inside of the walls due to rupture of the water tight walls is detected by the water detector, reactor scram can be conducted rapidly. (4) The sealing property of the flange joint portion is sufficient and detaching operation thereof is easy. (I.S.)

  17. 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.

  18. High pressure water jet mining machine

    Science.gov (United States)

    Barker, Clark R.

    1981-05-05

    A high pressure water jet mining machine for the longwall mining of coal is described. The machine is generally in the shape of a plowshare and is advanced in the direction in which the coal is cut. The machine has mounted thereon a plurality of nozzle modules each containing a high pressure water jet nozzle disposed to oscillate in a particular plane. The nozzle modules are oriented to cut in vertical and horizontal planes on the leading edge of the machine and the coal so cut is cleaved off by the wedge-shaped body.

  19. Water Delivery--It's All about Pressure

    Science.gov (United States)

    Roman, Harry T.

    2005-01-01

    There is a great deal of wisdom in the old saying "water seeks its level." In fact, the concept has bearing on a very practical side of human life as well, since the public water delivery system is based on it. In this article, the author discusses the concept behind water pressure and describes how the water systems work based on this concept.…

  20. Ultra-high pressure water jet: Baseline report

    International Nuclear Information System (INIS)

    1997-01-01

    The ultra-high pressure waterjet technology was being evaluated at Florida International University (FIU) as a baseline technology. In conjunction with FIU's evaluation of efficiency and cost, this report covers the evaluation conducted for safety and health issues. It is a commercially available technology and has been used for various projects at locations throughout the country. The ultra-high pressure waterjet technology acts as a cutting tool for the removal of surface substrates. The Husky trademark pump feeds water to a lance that directs the high pressure water at the surface to be removed. The safety and health evaluation during the testing demonstration focused on two main areas of exposure. These were dust and noise. The dust exposure was found to be minimal, which would be expected due to the wet environment inherent in the technology, but noise exposure was at a significant level. Further testing for noise is recommended because of the outdoor environment where the testing demonstration took place. In addition, other areas of concern found were arm-hand vibration, ergonomics, heat stress, tripping hazards, electrical hazards, lockout/tagout, fall hazards, slipping hazards, hazards associated with the high pressure water, and hazards associated with air pressure systems

  1. Performance Evaluation of Pressure Transducers for Water Impacts

    Science.gov (United States)

    Vassilakos, Gregory J.; Stegall, David E.; Treadway, Sean

    2012-01-01

    The Orion Multi-Purpose Crew Vehicle is being designed for water landings. In order to benchmark the ability of engineering tools to predict water landing loads, test programs are underway for scale model and full-scale water impacts. These test programs are predicated on the reliable measurement of impact pressure histories. Tests have been performed with a variety of pressure transducers from various manufacturers. Both piezoelectric and piezoresistive devices have been tested. Effects such as thermal shock, pinching of the transducer head, and flushness of the transducer mounting have been studied. Data acquisition issues such as sampling rate and anti-aliasing filtering also have been studied. The response of pressure transducers have been compared side-by-side on an impulse test rig and on a 20-inch diameter hemisphere dropped into a pool of water. The results have identified a range of viable configurations for pressure measurement dependent on the objectives of the test program.

  2. 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)

  3. 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

  4. 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

  5. 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

  6. Response of the water status of soybean to changes in soil water potentials controlled by the water pressure in microporous tubes

    Science.gov (United States)

    Steinberg, S. L.; Henninger, D. L.

    1997-01-01

    Water transport through a microporous tube-soil-plant system was investigated by measuring the response of soil and plant water status to step change reductions in the water pressure within the tubes. Soybeans were germinated and grown in a porous ceramic 'soil' at a porous tube water pressure of -0.5 kpa for 28 d. During this time, the soil matric potential was nearly in equilibrium with tube water pressure. Water pressure in the porous tubes was then reduced to either -1.0, -1.5 or -2.0 kPa. Sap flow rates, leaf conductance and soil, root and leaf water potentials were measured before and after this change. A reduction in porous tube water pressure from -0.5 to -1.0 or -1.5 kPa did not result in any significant change in soil or plant water status. A reduction in porous tube water pressure to -2.0 kPa resulted in significant reductions in sap flow, leaf conductance, and soil, root and leaf water potentials. Hydraulic conductance, calculated as the transpiration rate/delta psi between two points in the water transport pathway, was used to analyse water transport through the tube-soil-plant continuum. At porous tube water pressures of -0.5 to-1.5 kPa soil moisture was readily available and hydraulic conductance of the plant limited water transport. At -2.0 kPa, hydraulic conductance of the bulk soil was the dominant factor in water movement.

  7. 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.

  8. 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

  9. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T; Grunwald, G

    1998-10-01

    The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

  10. 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.

  11. Thermo-hydraulic behavior of saturated steam-water mixture in pressure vessel during injection of cold water

    International Nuclear Information System (INIS)

    Aya, Izuo; Kobayashi, Michiyuki; Inasaka, Fujio; Nariai, Hideki.

    1983-01-01

    The thermo-hydraulic behavior of saturated steam water mixture in a pressure vessel during injection of cold water was experimentally investigated with the Facility for Mixing Effect of Emergency Core Cooling Water. The dimensions of the pressure vessel used in the experiments were 284mm ID and 1,971mm height. 11 experiments were conducted without blowdown in order to comprehend the basic process excluding the effect of blowdown at injection of cold water. The initial pressure and water level, the injection flow rate and the size of injection nozzle were chosen as experimental parameters. Temperatures and void fractions at 6 elevations as well as pressure in the pressure vessel were measured, and new data especially on the pressure undershoot just after the initation of water injection and the vertical distribution of temperature and void fraction were gotten. The transients of pressure, average temperature and void fraction were caluculated using single-volume analysis code BLODAC-1V which is based on thermal equilibrium and so-called bubble gradient model. Some input parameters included in the analysis code were evaluated through the comparison of analysis with experimental data. Moreover, the observed pressure undershoot which is evaluated to be induced by a time lag of vapourization in water due to thermal nonequilibrium, was also discussed with the aid of another simple analysis model. (author)

  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. High-pressure water electrolysis: Electrochemical mitigation of product gas crossover

    International Nuclear Information System (INIS)

    Schalenbach, Maximilian; Stolten, Detlef

    2015-01-01

    Highlights: • New technique to reduce gas crossover during water electrolysis • Increase of the efficiency of pressurized water electrolysis • Prevention of safety hazards due to explosive gas mixtures caused by crossover • Experimental realization for a polymer electrolyte membrane electrolyzer • Discussion of electrochemical crossover mitigation for alkaline water electrolysis - Abstract: Hydrogen produced by water electrolysis can be used as an energy carrier storing electricity generated from renewables. During water electrolysis hydrogen can be evolved under pressure at isothermal conditions, enabling highly efficient compression. However, the permeation of hydrogen through the electrolyte increases with operating pressure and leads to efficiency loss and safety hazards. In this study, we report on an innovative concept, where the hydrogen crossover is electrochemically mitigated by an additional electrode between the anode and the cathode of the electrolysis cell. Experimentally, the technique was applied to a proton exchange membrane water electrolyzer operated at a hydrogen pressure that was fifty times larger than the oxygen pressure. Therewith, the hydrogen crossover was reduced and the current efficiency during partial load operation was increased. The concept is also discussed for water electrolysis that is operated at balanced pressures, where the crossover of hydrogen and oxygen is mitigated using two additional electrodes

  14. Experimental Study and Engineering Practice of Pressured Water Coupling Blasting

    Directory of Open Access Journals (Sweden)

    J. X. Yang

    2017-01-01

    Full Text Available Overburden strata movement in large space stope is the major reason that induces the appearance of strong mining pressure. Presplitting blasting for hard coal rocks is crucial for the prevention and control of strong pressure in stope. In this study, pressured water coupling blasting technique was proposed. The process and effect of blasting were analyzed by orthogonal test and field practice. Results showed that the presence of pressure-bearing water and explosive cartridges in the drill are the main influence factors of the blasting effect of cement test block. The high load-transmitting performance of pore water and energy accumulation in explosive cartridges were analyzed. Noxious substances produced during the blasting process were properly controlled because of the moistening, cooling, and diluting effect of pore water. Not only the goal of safe and static rock fragmentation by high-explosive detonation but also a combination of superdynamic blast loading and static loading effect of the pressured water was achieved. Then the practice of blasting control of hard coal rocks in Datong coal mine was analyzed to determine reasonable parameters of pressured water coupling blasting. A good presplitting blasting control effect was achieved for the hard coal rocks.

  15. 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.

  16. 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.

  17. Pressure dependence of dynamical heterogeneity in water

    International Nuclear Information System (INIS)

    Teboul, Victor

    2008-01-01

    Using molecular dynamics simulations we investigate the effect of pressure on the dynamical heterogeneity in water. We show that the effect of a pressure variation in water is qualitatively different from the effect of a temperature variation on the dynamical heterogeneity in the liquid. We observe a strong decrease of the aggregation of molecules of low mobility together with a decrease of the characteristic time associated with this aggregation. However, the aggregation of the most mobile molecules and the characteristic time of this aggregation are only slightly affected. In accordance with this result, the non-Gaussian parameter shows an important decrease with pressure while the characteristic time t* of the non-Gaussian parameter is only slightly affected. These results highlight then the importance of pressure variation investigations in low temperature liquids on approach to the glass transition

  18. 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

  19. Natural circulation in pressurized water reactors

    International Nuclear Information System (INIS)

    Bastos, J.L.F.; Loureiro, L.V.; Rocha, R.T.V. da; Umbehaun, P.E.

    1992-01-01

    Several analytical modelling have been done for steady-state and slow transients conditions, besides more sophisticated studies considering two and three dimensional effects in a very simple geometry. Under severe accident conditions for PWR a code to analyse natural circulation has been developed by Westinghouse. This paper discusses the problem of natural circulation in a complex geometry similar to that of nuclear power plants. A first experiment has been done at the integral test facility of 'Co-ordination of Special Projects-Ministry of Naval Affairs' (Coordenadoria para Projetos Especiais -Ministerio da Marinha, COPESP) for several flux conditions. The results obtained were compared with numerical simulations for the steady-state regime. 09 refs, 05 figs, 01 tab. (B.C.A.)

  20. 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

  1. The phase diagram of water at negative pressures: virtual ices.

    Science.gov (United States)

    Conde, M M; Vega, C; Tribello, G A; Slater, B

    2009-07-21

    The phase diagram of water at negative pressures as obtained from computer simulations for two models of water, TIP4P/2005 and TIP5P is presented. Several solid structures with lower densities than ice Ih, so-called virtual ices, were considered as possible candidates to occupy the negative pressure region of the phase diagram of water. In particular the empty hydrate structures sI, sII, and sH and another, recently proposed, low-density ice structure. The relative stabilities of these structures at 0 K was determined using empirical water potentials and density functional theory calculations. By performing free energy calculations and Gibbs-Duhem integration the phase diagram of TIP4P/2005 was determined at negative pressures. The empty hydrates sII and sH appear to be the stable solid phases of water at negative pressures. The phase boundary between ice Ih and sII clathrate occurs at moderate negative pressures, while at large negative pressures sH becomes the most stable phase. This behavior is in reasonable agreement with what is observed in density functional theory calculations.

  2. 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.

  3. Pressure: the politechnics of water supply in Mumbai.

    Science.gov (United States)

    Anand, Nikhil

    2011-01-01

    In Mumbai, most all residents are delivered their daily supply of water for a few hours every day, on a water supply schedule. Subject to a more precarious supply than the city's upper-class residents, the city's settlers have to consistently demand that their water come on “time” and with “pressure.” Taking pressure seriously as both a social and natural force, in this article I focus on the ways in which settlers mobilize the pressures of politics, pumps, and pipes to get water. I show how these practices not only allow settlers to live in the city, but also produce what I call hydraulic citizenship—a form of belonging to the city made by effective political and technical connections to the city's infrastructure. Yet, not all settlers are able to get water from the city water department. The outcomes of settlers' efforts to access water depend on a complex matrix of socionatural relations that settlers make with city engineers and their hydraulic infrastructure. I show how these arrangements describe and produce the cultural politics of water in Mumbai. By focusing on the ways in which residents in a predominantly Muslim settlement draw water despite the state's neglect, I conclude by pointing to the indeterminacy of water, and the ways in which its seepage and leakage make different kinds of politics and publics possible in the city.

  4. Water hammer characteristics of integral pressurized water reactor primary loop

    International Nuclear Information System (INIS)

    Zuo, Qiaolin; Qiu, Suizheng; Lu, Wei; Tian, Wenxi; Su, Guanghui; Xiao, Zejun

    2013-01-01

    Highlights: • Water hammer models developed for IPWR primary loop using MOC. • Good agreement between the developed code and the experiment. • The good agreement between WAHAP and Flowmaster can validate the equations in WAHAP. • The primary loop of IPWR suffers from slight water hammer impact. -- Abstract: The present work discussed the single-phase water hammer phenomenon, which was caused by the four-pump-alternate startup in an integral pressurized water reactor (IPWR). A new code named water hammer program (WAHAP) was developed independently based on the method of characteristic to simulate hydraulic transients in the primary system of IPWR and its components such as reactor core, once-through steam generators (OTSG), the main coolant pumps and so on. Experimental validation for the correctness of the equations and models in WAHAP was carried out and the models fit the experimental data well. Some important variables were monitored including transient volume flow rates, opening angle of valve disc and pressure drop in valves. The water hammer commercial software Flowmaster V7 was also employed to compare with WAHAP and the good agreement can validate the equations in WAHAP. The transient results indicated that the primary loop of IPWR suffers from slight water hammer impact under pump switching conditions

  5. Water hammer characteristics of integral pressurized water reactor primary loop

    Energy Technology Data Exchange (ETDEWEB)

    Zuo, Qiaolin [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Qiu, Suizheng, E-mail: szqiu@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Lu, Wei; Tian, Wenxi; Su, Guanghui [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Xiao, Zejun [Nuclear Power Institute of China, Chengdu, Sichuan 610041 (China)

    2013-08-15

    Highlights: • Water hammer models developed for IPWR primary loop using MOC. • Good agreement between the developed code and the experiment. • The good agreement between WAHAP and Flowmaster can validate the equations in WAHAP. • The primary loop of IPWR suffers from slight water hammer impact. -- Abstract: The present work discussed the single-phase water hammer phenomenon, which was caused by the four-pump-alternate startup in an integral pressurized water reactor (IPWR). A new code named water hammer program (WAHAP) was developed independently based on the method of characteristic to simulate hydraulic transients in the primary system of IPWR and its components such as reactor core, once-through steam generators (OTSG), the main coolant pumps and so on. Experimental validation for the correctness of the equations and models in WAHAP was carried out and the models fit the experimental data well. Some important variables were monitored including transient volume flow rates, opening angle of valve disc and pressure drop in valves. The water hammer commercial software Flowmaster V7 was also employed to compare with WAHAP and the good agreement can validate the equations in WAHAP. The transient results indicated that the primary loop of IPWR suffers from slight water hammer impact under pump switching conditions.

  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. Pressurized water reactor fuel rod design methodology

    International Nuclear Information System (INIS)

    Silva, A.T.; Esteves, A.M.

    1988-08-01

    The fuel performance program FRAPCON-1 and the structural finite element program SAP-IV are applied in a pressurized water reactor fuel rod design methodology. The applied calculation procedure allows to dimension the fuel rod components and characterize its internal pressure. (author) [pt

  8. In Situ Raman Study of Liquid Water at High Pressure.

    Science.gov (United States)

    Romanenko, Alexandr V; Rashchenko, Sergey V; Goryainov, Sergey V; Likhacheva, Anna Yu; Korsakov, Andrey V

    2018-06-01

    A pressure shift of Raman band of liquid water (H 2 O) may be an important tool for measuring residual pressures in mineral inclusions, in situ barometry in high-pressure cells, and as an indicator of pressure-induced structural transitions in H 2 O. However, there was no consensus as to how the broad and asymmetric water Raman band should be quantitatively described, which has led to fundamental inconsistencies between reported data. In order to overcome this issue, we measured Raman spectra of H 2 O in situ up to 1.2 GPa using a diamond anvil cell, and use them to test different approaches proposed for the description of the water Raman band. We found that the most physically meaningful description of water Raman band is the decomposition into a linear background and three Gaussian components, associated with differently H-bonded H 2 O molecules. Two of these components demonstrate a pronounced anomaly in pressure shift near 0.4 GPa, supporting ideas of structural transition in H 2 O at this pressure. The most convenient approach for pressure calibration is the use of "a linear background + one Gaussian" decomposition (the pressure can be measured using the formula P (GPa) = -0.0317(3)·Δν G (cm -1 ), where Δν G represents the difference between the position of water Raman band, fitted as a single Gaussian, in measured spectrum and spectrum at ambient pressure).

  9. A review on water fault diagnosis of PEMFC associated with the pressure drop

    International Nuclear Information System (INIS)

    Pei, Pucheng; Li, Yuehua; Xu, Huachi; Wu, Ziyao

    2016-01-01

    Highlights: • Reviewed the effect factors and estimations of pressure drop associated with water fault diagnosis. • Reviewed pressure drop-based water fault diagnosis using different indicators. • Deviation of pressure drop is used frequently to diagnose water fault. • Reviewed recovery strategies based on pressure drop used in commercial PEMFC. • Merits, demerits and application prospects of pressure drop-based water fault diagnosis are discussed. - Abstract: The pressure difference between the inlet and outlet of the reactant in fuel cells is called the pressure drop, which is related to the water amount inside the fuel cells. In recent years there have been many studies that used the pressure drop to detect the water content and diagnose water fault of proton exchange membrane fuel cells (PEMFCs). To our knowledge, there has not been a systematic review of these studies. In this paper, the effect variables of pressure drop are reviewed firstly. Then estimations of the theoretical pressure drop are reviewed mainly based on the following four aspects: Bernoulli’s equation, two-phase flow multiplier, Darcy’s law and artificial intelligence. Afterward, the water fault diagnosis based on the pressure drop using the following six indicators are reviewed: indicator of direct pressure drop, its deviation, frequency, multiplier, the ratio of pressure drop to flow rate and the flooding degree. In addition, the strategies of water fault recovery are also summarized. Finally the merits, demerits and application prospects of pressure drop-based water fault diagnosis are presented.

  10. Supercritical-pressure light water cooled reactors

    CERN Document Server

    Oka, Yoshiaki

    2014-01-01

    This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

  11. SIMULATION OF NEGATIVE PRESSURE WAVE PROPAGATION IN WATER PIPE NETWORK

    Directory of Open Access Journals (Sweden)

    Tang Van Lam

    2017-11-01

    Full Text Available Subject: factors such as pipe wall roughness, mechanical properties of pipe materials, physical properties of water affect the pressure surge in the water supply pipes. These factors make it difficult to analyze the transient problem of pressure evolution using simple programming language, especially in the studies that consider only the magnitude of the positive pressure surge with the negative pressure phase being neglected. Research objectives: determine the magnitude of the negative pressure in the pipes on the experimental model. The propagation distance of the negative pressure wave will be simulated by the valve closure scenarios with the help of the HAMMER software and it is compared with an experimental model to verify the quality the results. Materials and methods: academic version of the Bentley HAMMER software is used to simulate the pressure surge wave propagation due to closure of the valve in water supply pipe network. The method of characteristics is used to solve the governing equations of transient process of pressure change in the pipeline. This method is implemented in the HAMMER software to calculate the pressure surge value in the pipes. Results: the method has been applied for water pipe networks of experimental model, the results show the affected area of negative pressure wave from valve closure and thereby we assess the largest negative pressure that may appear in water supply pipes. Conclusions: the experiment simulates the water pipe network with a consumption node for various valve closure scenarios to determine possibility of appearance of maximum negative pressure value in the pipes. Determination of these values in real-life network is relatively costly and time-consuming but nevertheless necessary for identification of the risk of pipe failure, and therefore, this paper proposes using the simulation model by the HAMMER software. Initial calibration of the model combined with the software simulation results and

  12. Neutron fluence determination for light water reactor pressure vessels

    International Nuclear Information System (INIS)

    Gold, R.

    1994-01-01

    A general description of limitations that exist in pressure vessel neutron fluence determinations for commercial light water reactors is presented. Complexity factors that arise in light water reactor pressure vessel neutron fluence calculations are identified and used to analyze calculational limitations. Two broad categories of calculational limitations are introduced, namely benchmark field limitations and deep penetration limitations. Explicit examples of limitations that can arise in each of these two broad categories are presented. These limitations are used to show that the recent draft regulatory guide for the determination of pressure vessel neutron fluence, developed by the Nuclear Regulatory Commission, is based upon procedures and assumptions that are not valid. To eliminate the complexity and limitations of calculational methods, it is recommended that the determination of light water reactor pressure vessel neutron fluence be based upon experiment. Recommendations for improved methods of pressure vessel surveillance neutron dosimetry are advanced

  13. Is high-pressure water the cradle of life?

    International Nuclear Information System (INIS)

    Bassez, Marie-Paule

    2003-01-01

    Several theories have been proposed for the synthesis of prebiotic molecules. This letter shows that the structure of supercritical water, or high-pressure water, could trigger prebiotic synthesis and the origin of life deep in the oceans, in hydrothermal vent systems. Dimer geometries of high-pressure water may have a point of symmetry and a zero dipole moment. Consequently, simple apolar molecules found in submarine hydrothermal vent systems will dissolve in the apolar environment provided by the apolar form of the water dimer. Apolar water could be the medium which helps precursor molecules to concentrate and react more efficiently. The formation of prebiotic molecules could thus be linked to the structure of the water inside chimney nanochannels and cavities where hydrothermal piezochemistry and shock wave chemistry could occur. (letter to the editor)

  14. Capillarity Induced Negative Pressure of Water Plugs in Nanochannels

    NARCIS (Netherlands)

    Tas, Niels Roelof; Mela, P.; Kramer, Tobias; Berenschot, Johan W.; van den Berg, Albert

    2003-01-01

    We have found evidence that water plugs in hydrophilic nanochannels can be at significant negative pressure due to tensile capillary forces. The negative pressure of water plugs in nanochannels induces bending of the thin channel capping layer, which results in a visible curvature of the liquid

  15. AP600 design certification thermal hydraulics testing and analysis

    Energy Technology Data Exchange (ETDEWEB)

    Hochreiter, L.E.; Piplica, E.J.

    1995-09-01

    Westinghouse Electric Corporation, in conjunction with the Department of Energy and the Electric Power Research Institute, have been developing an advanced light water reactor design; the AP600. The AP600 is a 1940 Mwt, 600Mwe unit which is similar to a Westinghouse two-loop Pressurized Water Reactor. The accumulated knowledge on reactor design to reduce the capital costs, construction time, and the operational and maintenance cost of the unit once it begins to generate electrical power. The AP600 design goal is to maintain an overall cost advantage over fossil generated electrical power.

  16. 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

  17. 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

  18. 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

  19. Reactor water spontaneous circulation structure in reactor pressure vessel

    International Nuclear Information System (INIS)

    Takahashi, Kazumi

    1998-01-01

    The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

  20. Reactor vessel pressure transient protection for pressurized water reactors

    International Nuclear Information System (INIS)

    Zech, G.

    1978-09-01

    During the past few years the NRC has been studying the issue of protection of the reactor pressure vessels at Pressurized Water Reactors (PWRs) from transients when the vessels are at a relatively low temperature. This effort was prompted by concerns related to the safety margins available to vessel damage as a result of such events. Nuclear Reactor Regulation Category A Technical Activity No. A-26 was established to set forth the NRC plan for resolution of the generic aspects of this safety issue. The purpose of the report is to document the completion of this generic technical activity

  1. 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.)

  2. Tensile Strength of Water Exposed to Pressure Pulses

    DEFF Research Database (Denmark)

    Andersen, Anders Peter; Mørch, Knud Aage

    2012-01-01

    at an extended water-solid interface by imposing a tensile stress pulse which easily causes cavitation. Next, a compressive pulse of duration ~1 ms and a peak intensity of a few bar is imposed prior to the tensile stress pulse. A dramatic increase of the tensile strength is observed immediately after......It is well known that pressurization for an extended period of time increases the tensile strength of water, but little information is available on the effect of pressure pulses of short duration. This is addressed in the present paper where we first measure the tensile strength of water...

  3. Analytical study on water hammer pressure in pressurized conduits with a throttled surge chamber for slow closure

    Directory of Open Access Journals (Sweden)

    Yong-liang Zhang

    2010-06-01

    Full Text Available This paper presents an analytical investigation of water hammer in a hydraulic pressurized pipe system with a throttled surge chamber located at the junction between a tunnel and a penstock, and a valve positioned at the downstream end of the penstock. Analytical formulas of maximum water hammer pressures at the downstream end of the tunnel and the valve were derived for a system subjected to linear and slow valve closure. The analytical results were then compared with numerical ones obtained using the method of characteristics. There is agreement between them. The formulas can be applied to estimating water hammer pressure at the valve and transmission of water hammer pressure through the surge chamber at the junction for a hydraulic pipe system with a surge chamber.

  4. 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

  5. High pressure water jet cutting and stripping

    Science.gov (United States)

    Hoppe, David T.; Babai, Majid K.

    1991-01-01

    High pressure water cutting techniques have a wide range of applications to the American space effort. Hydroblasting techniques are commonly used during the refurbishment of the reusable solid rocket motors. The process can be controlled to strip a thermal protective ablator without incurring any damage to the painted surface underneath by using a variation of possible parameters. Hydroblasting is a technique which is easily automated. Automation removes personnel from the hostile environment of the high pressure water. Computer controlled robots can perform the same task in a fraction of the time that would be required by manual operation.

  6. Decontamination and recycle of zirconium pressure tubes from Pressurized Heavy Water Reactor

    International Nuclear Information System (INIS)

    Gantayet, L.M.; Verma, R.; Remya Devi, P.S.; Banerjee, S.; Kotak, V.; Raha, A.; Sandeep, K.C.; Joshi, Shreeram W.; Lali, A.M.

    2009-01-01

    An ion exchange process has been developed for decontamination of zirconium pressure tubes from Pressurized Heavy Water Reactor and recycling of neutronically improved zirconium. Distribution coefficient, equilibrium isotherm, kinetic and breakthrough data were used to develop the separation process. Effect of gamma radiation on indigenous resins was also studied to assess their suitability in high radiation field. (author)

  7. Structure and dynamics of water confined in a graphene nanochannel under gigapascal high pressure: dependence of friction on pressure and confinement.

    Science.gov (United States)

    Yang, Lei; Guo, Yanjie; Diao, Dongfeng

    2017-05-31

    Recently, water flow confined in nanochannels has become an interesting topic due to its unique properties and potential applications in nanofluidic devices. The trapped water is predicted to experience high pressure in the gigapascal regime. Theoretical and experimental studies have reported various novel structures of the confined water under high pressure. However, the role of this high pressure on the dynamic properties of water has not been elucidated to date. In the present study, the structure evolution and interfacial friction behavior of water constrained in a graphene nanochannel were investigated via molecular dynamics simulations. Transitions of the confined water to different ice phases at room temperature were observed in the presence of lateral pressure at the gigapascal level. The friction coefficient at the water/graphene interface was found to be dependent on the lateral pressure and nanochannel height. Further theoretical analyses indicate that the pressure dependence of friction is related to the pressure-induced change in the structure of water and the confinement dependence results from the variation in the water/graphene interaction energy barrier. These findings provide a basic understanding of the dynamics of the nanoconfined water, which is crucial in both fundamental and applied science.

  8. 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

  9. 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.)

  10. 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

  11. 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

  12. 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

  13. Cavitation nuclei in water exposed to transient pressures

    DEFF Research Database (Denmark)

    Andersen, Anders Peter; Mørch, Knud Aage

    2015-01-01

    A model of skin-stabilized interfacial cavitation nuclei and their response to tensile and compressive stressing is presented. The model is evaluated in relation to experimental tensile strength results for water at rest at the bottom of an open water-filled container at atmospheric pressure...... and room temperature. These results are obtained by recording the initial growth of cavities generated by a short tensile pulse applied to the bottom of the container. It is found that the cavitation nuclei shift their tensile strength depending on their pressure history. Static pressurization...... for an extended period of time prior to testing is known to increase the tensile strength of water, but little information is available on how it is affected by compression pulses of short duration. This is addressed by imposing compression pulses of approximately 1 ms duration and a peak intensity of a few bar...

  14. Water cycles in closed ecological systems: effects of atmospheric pressure.

    Science.gov (United States)

    Rygalov, Vadim Y; Fowler, Philip A; Metz, Joannah M; Wheeler, Raymond M; Bucklin, Ray A

    2002-01-01

    In bioregenerative life support systems that use plants to generate food and oxygen, the largest mass flux between the plants and their surrounding environment will be water. This water cycle is a consequence of the continuous change of state (evaporation-condensation) from liquid to gas through the process of transpiration and the need to transfer heat (cool) and dehumidify the plant growth chamber. Evapotranspiration rates for full plant canopies can range from ~1 to 10 L m-2 d-1 (~1 to 10 mm m-2 d-1), with the rates depending primarily on the vapor pressure deficit (VPD) between the leaves and the air inside the plant growth chamber. VPD in turn is dependent on the air temperature, leaf temperature, and current value of relative humidity (RH). Concepts for developing closed plant growth systems, such as greenhouses for Mars, have been discussed for many years and the feasibility of such systems will depend on the overall system costs and reliability. One approach for reducing system costs would be to reduce the operating pressure within the greenhouse to reduce structural mass and gas leakage. But managing plant growth environments at low pressures (e.g., controlling humidity and heat exchange) may be difficult, and the effects of low-pressure environments on plant growth and system water cycling need further study. We present experimental evidence to show that water saturation pressures in air under isothermal conditions are only slightly affected by total pressure, but the overall water flux from evaporating surfaces can increase as pressure decreases. Mathematical models describing these observations are presented, along with discussion of the importance for considering "water cycles" in closed bioregenerative life support systems.

  15. Water cycles in closed ecological systems: effects of atmospheric pressure

    Science.gov (United States)

    Rygalov, Vadim Y.; Fowler, Philip A.; Metz, Joannah M.; Wheeler, Raymond M.; Bucklin, Ray A.; Sager, J. C. (Principal Investigator)

    2002-01-01

    In bioregenerative life support systems that use plants to generate food and oxygen, the largest mass flux between the plants and their surrounding environment will be water. This water cycle is a consequence of the continuous change of state (evaporation-condensation) from liquid to gas through the process of transpiration and the need to transfer heat (cool) and dehumidify the plant growth chamber. Evapotranspiration rates for full plant canopies can range from 1 to 10 L m-2 d-1 (1 to 10 mm m-2 d-1), with the rates depending primarily on the vapor pressure deficit (VPD) between the leaves and the air inside the plant growth chamber. VPD in turn is dependent on the air temperature, leaf temperature, and current value of relative humidity (RH). Concepts for developing closed plant growth systems, such as greenhouses for Mars, have been discussed for many years and the feasibility of such systems will depend on the overall system costs and reliability. One approach for reducing system costs would be to reduce the operating pressure within the greenhouse to reduce structural mass and gas leakage. But managing plant growth environments at low pressures (e.g., controlling humidity and heat exchange) may be difficult, and the effects of low-pressure environments on plant growth and system water cycling need further study. We present experimental evidence to show that water saturation pressures in air under isothermal conditions are only slightly affected by total pressure, but the overall water flux from evaporating surfaces can increase as pressure decreases. Mathematical models describing these observations are presented, along with discussion of the importance for considering "water cycles" in closed bioregenerative life support systems.

  16. Pressurized water reactor inspection procedures

    International Nuclear Information System (INIS)

    Heinrich, D.; Mueller, G.; Otte, H.J.; Roth, W.

    1998-01-01

    Inspections of the reactor pressure vessels of pressurized water reactors (PWR) so far used to be carried out with different central mast manipulators. For technical reasons, parallel inspections of two manipulators alongside work on the refueling cavity, so as to reduce the time spent on the critical path in a revision outage, are not possible. Efforts made to minimize the inspection time required with one manipulator have been successful, but their effects are limited. Major reductions in inspection time can be achieved only if inspections are run with two manipulators in parallel. The decentralized manipulator built by GEC Alsthom Energie and so far emmployed in boiling water reactors in the USA, Spain, Switzerland and Japan allows two systems to be used in parallel, thus reducing the time required for standard inspection of a pressure vessel from some six days to three days. These savings of approximately three days are made possible without any compromises in terms of positioning by rail-bound systems. During inspection, the reactor refueling cavity is available for other revision work without any restrictions. The manipulator can be used equally well for inspecting standard PWR, PWR with a thermal shield, for inspecting the land between in-core instrumentation nozzles, BWR with and without jet pumps (complementary inspection), and for inspecting core support shrouds. (orig.) [de

  17. 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)

  18. Determination of a test section parameters for Iris nuclear reactor pressurizer

    International Nuclear Information System (INIS)

    Silva, Mario A.B. da; Lira, Carlos A.B. de O.

    2009-01-01

    An integral, modular and medium size nuclear reactor, known as IRIS, is being developed by Westinghouse and by research centers. IRIS is characterized by having most of its components inside the pressure vessel, eliminating the probability of accidents. Due to its integral configuration, there is no spray system for boron homogenization, which may cause power transients. Thus, boron mixing must be investigated. The aim of this paper is to establish the conditions under which a test section has to be built for boron dispersion analysis inside IRIS reactor pressurizer. Through Fractional Scaling Analysis, which is a new methodology of similarity, the main parameters for a test section are obtained. By combining Fractional Scaling Analysis with local scaling for the densimetric Froude number and a previously established volumetric scale factor, the values of recirculation orifices, inlet water temperature, time scale factor and recirculation flow for the test section (model) are determined so that boron distribution is well represented in IRIS reactor pressurizer (prototype). Analytical solutions were used to validate the adopted methodology and when the results simulated in the model are compared to those that characterize the prototype, the agreement for both systems is absolute. The thermal power also influences boron distribution inside the test section. This power is determined by condensation laws in the vapor region and by suitable correlations for free convection. The fractions for rising inlet recirculation water enthalpy and vapor formation are also considered. (author)

  19. Variation of Pore Water Pressure in Tailing Sand under Dynamic Loading

    Directory of Open Access Journals (Sweden)

    Jia-xu Jin

    2018-01-01

    Full Text Available Intense vibration affects the pore water pressure in a tailing dam, with the tendency to induce dam liquefaction. In this study, experiments were performed wherein model tailing dams were completely liquefied by sustained horizontal dynamic loading to determine the effects of the vibration frequency, vibration amplitude, and tailing density on the pore water pressure. The results revealed four stages in the increase of the tailing pore water pressure under dynamic loading, namely, a slow increase, a rapid increase, inducement of structural failure, and inducement of complete liquefaction. A lower frequency and smaller amplitude of the vibration were found to increase the time required to achieve a given pore water pressure in dense tailings. Under the effect of these three factors—vibration frequency and amplitude and tailing density—the tailing liquefaction time varied nonlinearly with the height from the base of the tailing dam, with an initial decrease followed by an increase. The pore pressure that induced structural failure also gradually decreased with increasing height. The increase in the tailing pore pressure could be described by an S-shaped model. A complementary multivariate nonlinear equation was also derived for predicting the tailing pore water pressure under dynamic loading.

  20. The pressurized water reactor

    International Nuclear Information System (INIS)

    Gallagher, J.L.

    1987-01-01

    Pressurized water reactor technology has reached a maturity that has engendered a new surge of innovation, which in turn, has led to significant advances in the technology. These advances, characterized by bold thinking but conservative execution, are resulting in nuclear plant designs which offer significant performance and safety improvements. This paper describes the innovations which are being designed into mainstream PWR technology as well as the desings which are resulting from such innovations. (author)

  1. Stability analysis of supercritical-pressure light water-cooled reactor in constant pressure operation

    International Nuclear Information System (INIS)

    Suhwan, JI; Shirahama, H.; Koshizuka, S.; Oka, Y.

    2001-01-01

    The purpose of this study is to evaluate the thermal-hydraulic and the thermal-nuclear coupled stabilities of a supercritical pressure light water-cooled reactor. A stability analysis code at supercritical pressure is developed. Using this code, stabilities of full and partial-power reactor operating at supercritical pressure are investigated by the frequency-domain analysis. Two types of SCRs are analyzed; a supercritical light water reactor (SCLWR) and a supercritical water-cooled fast reactor (SCFR). The same stability criteria as Boiling Water Reactor are applied. The thermal-hydraulic stability of SCLWR and SCFR satisfies the criteria with a reasonable orifice loss coefficient. The decay ratio of the thermal-nuclear coupled stability in SCFR is almost zero because of a small coolant density coefficient of the fast reactor. The evaluated decay ratio of the thermal-nuclear coupled stability is 3,41 ∼ 10 -V at 100% power in SCFR and 0,028 at 100% power in SCLWR. The sensitivity is investigated. It is found that the thermal-hydraulic stability is sensitive to the mass flow rate strongly and the thermal-nuclear coupled stability to the coolant density coefficient. The bottom power peak distribution makes the thermal-nuclear stability worse and the thermal-nuclear stability better. (author)

  2. Study on low pressure evaporation of fresh water generation system model

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Han Shik; Wibowo, Supriyanto; Shin, Yong Han; Jeong, Hyo Min [Gyeongsang National University, Tongyeong (Korea, Republic of); Fajar, Berkah [University of Diponegoro, Semarang (Indonesia)

    2012-02-15

    A low pressure evaporation fresh water generation system is designed for converting brackish water or seawater into fresh water by distillation in low pressure and temperature. Distillation through evaporation of feed water and subsequent vapor condensation as evaporation produced fresh water were studied; tap water was employed as feed water. The system uses the ejector as a vacuum creator of the evaporator, which is one of the most important parts in the distillation process. Hence liquid can be evaporated at a lower temperature than at normal or atmospheric conditions. Various operating conditions, i.e. temperature of feed water and different orifice diameters, were applied in the experiment to investigate the characteristics of the system. It was found that these parameters have a significant effect on the performance of fresh water generation systems with low pressure evaporation.

  3. Nonlinear vibration of a hemispherical dome under external water pressure

    International Nuclear Information System (INIS)

    Ross, C T F; McLennan, A; Little, A P F

    2011-01-01

    The aim of this study was to analyse the behaviour of a hemi-spherical dome when vibrated under external water pressure, using the commercial computer package ANSYS 11.0. In order to achieve this aim, the dome was modelled and vibrated in air and then in water, before finally being vibrated under external water pressure. The results collected during each of the analyses were compared to the previous studies, and this demonstrated that ANSYS was a suitable program and produced accurate results for this type of analysis, together with excellent graphical displays. The analysis under external water pressure, clearly demonstrated that as external water pressure was increased, the resonant frequencies decreased and a type of dynamic buckling became likely; because the static buckling eigenmode was similar to the vibration eigenmode. ANSYS compared favourably with the in-house software, but had the advantage that it produced graphical displays. This also led to the identification of previously undetected meridional modes of vibration; which were not detected with the in-house software.

  4. Nonlinear vibration of a hemispherical dome under external water pressure

    Science.gov (United States)

    Ross, C. T. F.; McLennan, A.; Little, A. P. F.

    2011-07-01

    The aim of this study was to analyse the behaviour of a hemi-spherical dome when vibrated under external water pressure, using the commercial computer package ANSYS 11.0. In order to achieve this aim, the dome was modelled and vibrated in air and then in water, before finally being vibrated under external water pressure. The results collected during each of the analyses were compared to the previous studies, and this demonstrated that ANSYS was a suitable program and produced accurate results for this type of analysis, together with excellent graphical displays. The analysis under external water pressure, clearly demonstrated that as external water pressure was increased, the resonant frequencies decreased and a type of dynamic buckling became likely; because the static buckling eigenmode was similar to the vibration eigenmode. ANSYS compared favourably with the in-house software, but had the advantage that it produced graphical displays. This also led to the identification of previously undetected meridional modes of vibration; which were not detected with the in-house software.

  5. When immiscible becomes miscible-Methane in water at high pressures.

    Science.gov (United States)

    Pruteanu, Ciprian G; Ackland, Graeme J; Poon, Wilson C K; Loveday, John S

    2017-08-01

    At low pressures, the solubility of gases in liquids is governed by Henry's law, which states that the saturated solubility of a gas in a liquid is proportional to the partial pressure of the gas. As the pressure increases, most gases depart from this ideal behavior in a sublinear fashion, leveling off at pressures in the 1- to 5-kbar (0.1 to 0.5 GPa) range with solubilities of less than 1 mole percent (mol %). This contrasts strikingly with the well-known marked increase in solubility of simple gases in water at high temperature associated with the critical point (647 K and 212 bar). The solubility of the smallest hydrocarbon, the simple gas methane, in water under a range of pressure and temperature is of widespread importance, because it is a paradigmatic hydrophobe and occurs widely in terrestrial and extraterrestrial geology. We report measurements up to 3.5 GPa of the pressure dependence of the solubility of methane in water at 100°C-well below the latter's critical temperature. Our results reveal a marked increase in solubility between 1 and 2 GPa, leading to a state above 2 GPa where the maximum solubility of methane in water exceeds 35 mol %.

  6. Flooding of a large, passive, pressure-tube light water reactor

    International Nuclear Information System (INIS)

    Hejzlar, P.; Todreas, N.E.; Driscoll, M.J.

    1997-01-01

    A reactor concept has been developed which can survive loss of coolant accidents without scram and without replenishing primary coolant inventory, while maintaining safe temperature limits on the fuel and pressure tubes. The proposed concept is a pressure tube type reactor of similar design to CANDU reactors, but differing in three key aspects. First, a solid SiC-coated graphite fuel matrix is used in place of fuel pin bundles to enable the dissipation of decay heat from the fuel in the absence of primary coolant. Second, the heavy water coolant in the pressure tubes is replaced by light water, which also serves as the moderator. Finally, the calandria tank, surrounded by a graphite reflector, contains a low pressure gas instead of heavy water moderator, and this normally-voided calandria is connected to a light water heat sink. The cover gas displaces the light water from the calandria during normal operation, while during loss of coolant or loss of heat sink accidents it allows passive calandria flooding. Calandria flooding also provides redundant and diverse reactor shutdown. This paper describes the thermal hydraulic characteristics of the passively initiated, gravity driven calandria flooding process. Flooding the calandria space with light water is a unique and very important feature of the proposed pressure-tube light water reactor (PTLWR) concept. The flooding of the top row of fuel channels must be accomplished fast enough so that in the total loss of coolant, none of the critical components of the fuel channel, i.e. the pressure tube, the calandria tube, the matrix and the fuel, exceed their design limits. The flooding process has been modeled and shown to be rapid enough to maintain all components within their design limits. (orig.)

  7. Pumps for German pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dernedde, R.

    1984-01-01

    The article describes the development of a selection of pumps which are used in the primary coolant system and the high-pressure safety injection system and feed water system during the past 2 decades. The modifications were caused by the step-wise increasing power output of the plants from 300 MW up to 1300 MW. Additional important influences were given be the increased requirements for quality assurance and final-documentation. The good operating results of the delivered pumps proved that the reliability is independent of the volume of the software-package. The outlook expects that consolidation will be followed by additional steps for the order processing of components for the convoy pumps. KW: main coolant pump; primary system; boiler feed pump; reactor pump; secondary system; barrel insert pump; pressure water reactor; convoy pump; state of the art.

  8. Research on axial total pressure distributions of sonic steam jet in subcooled water

    International Nuclear Information System (INIS)

    Wu Xinzhuang; Li Wenjun; Yan Junjie

    2012-01-01

    The axial total pressure distributions of sonic steam jet in subcooled water were experimentally investigated for three different nozzle diameters (6.0 mm, 8.0 mm and 10.0 mm). The inlet steam pressure, and pool subcooling subcooled water temperature were in the range of 0.2-0.6 MPa and 420-860 ℃, respectively. The effect of steam pressure, subcooling water temperature and nozzle size on the axial pressure distributions were obtained, and also the characteristics of the maximum pressure and its position were studied. The results indicated that the characteristics of the maximum pressure were influenced by the nozzle size for low steam pressure, but the influence could be ignored for high steam pressure. Moreover, a correlation was given to correlate the position of the maximum pressure based on steam pressure and subcooling water temperature, and the discrepancies of predictions and experiments are within ±15%. (authors)

  9. The effect of pressurizer-water-level on the low frequency component of the pressure spectrum in a PWR

    International Nuclear Information System (INIS)

    Por, G.; Izsak, E.; Valko, J.

    1984-09-01

    The pressure fluctuations were measured in the cooling system of the Paks-1 reactor. A shift of the peak was detected in low frequency component of the pressure fluctuation spectrum which is due to the fluctuations of water level in the pressurizer. Using the model of Katona and Nagy (1983), the eigenfrequencies, the magnitude of the shift and the sensitivity to the pressurizer water level were reproduced in good accordance with the experimental data. (D.Gy.)

  10. 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

  11. Type GQS-1 high pressure steam manifold water level monitoring system

    International Nuclear Information System (INIS)

    Li Nianzu; Li Beicheng; Jia Shengming

    1993-10-01

    The GQS-1 high pressure steam manifold water level monitoring system is an advanced nuclear gauge that is suitable for on-line detecting and monitor in high pressure steam manifold water level. The physical variable of water level is transformed into electrical pulses by the nuclear sensor. A computer is equipped for data acquisition, analysis and processing and the results are displayed on a 14 inch color monitor. In addition, a 4 ∼ 20 mA output current is used for the recording and regulation of water level. The main application of this gauge is for on-line measurement of high pressure steam manifold water level in fossil-fired power plant and other industries

  12. Evaluation of pressure transducers under turbid natural waters

    Digital Repository Service at National Institute of Oceanography (India)

    Joseph, A.; Desa, E.; Desa, E.; Smith, D.; Peshwe, V.B.; VijayKumar, K.; Desa, J.A.E.

    , the use of rho sub(eff) in contrast to the bulk density, significantly improves the measurement accuracy. For celar waters, precision density measurements made on discrete water samples agreed with rho sub(eff) values derived from pressure measurements...

  13. Water pressure and ground vibrations induced by water guns near Bandon Road Lock and Dam and Lemont, Illinois

    Science.gov (United States)

    Adams, Ryan F.; Koebel, Carolyn M.; Morrow, William S.

    2018-02-13

    Multiple geophysical sensors were used to characterize the underwater pressure field and ground vibrations of a seismic water gun and its suitability to deter the movement of Asian carps (particularly the silver [Hypophthalmichthys molitrix] and bighead [Hypophthalmichthys nobilis] carps) while ensuring the integrity of surrounding structures. The sensors used to collect this information were blast-rated hydrophones, surface- and borehole-mounted geophones, and fixed accelerometers.Results from two separate studies are discussed in this report. The Brandon Road study took place in May 2014, in the Des Plaines River, in a concrete-walled channel downstream of the Brandon Road Lock and Dam near Joliet, Illinois. The Lemont study took place in June 2014, in a segment of the dolomite setblock-lined Chicago Sanitary and Ship Canal near Lemont, Illinois.Two criteria were evaluated to assess the potential deterrence to carp migration, and to minimize the expected effect on nearby structures from discharge of the seismic water gun. The first criterion was a 5-pound-per-square-inch (lb/in2) limit for dynamic underwater pressure variations. The second criterion was a maximum velocity and acceleration disturbance of 0.75 inch per second (in/s) for sensitive machinery (such as the lock gates and pumps) and 2.0 in/s adjacent to canal walls, respectively. The criteria were based on previous studies of fish responses to dynamic pressure variations, and effects of vibrations on the structural integrity of concrete walls.The Brandon Road study evaluated the magnitude and extent of the pressure field created by two water gun configurations in the concrete-walled channel downstream of the lock where channel depths ranged from 11 to 14 feet (ft). Data from a single 80-cubic-inch (in³) water gun set at 6 ft below water surface (bws) produced a roughly cylindrical 5-lb/in2 pressure field 20 ft in radius, oriented vertically, with the radius decreasing to less than 15 ft at the water

  14. Chloride Ingress into Concrete under Water Pressure

    DEFF Research Database (Denmark)

    Lund, Mia Schou; Sander, Lotte Braad; Grelk, Bent

    2011-01-01

    The chloride ingress into concrete under water pressures of 100 kPa and 800 kPa have been investigated by experiments. The specimens were exposed to a 10% NaCl solution and water mixture. For the concrete having w/c = 0.35 the experimental results show the chloride diffusion coefficient at 800 k......Pa (~8 atm.) is 12 times greater than at 100 kPa (~1 atm.). For w/c = 0.45 and w/c = 0.55 the chloride diffusion coefficients are 7 and 3 times greater. This means that a change in pressure highly influences the chloride ingress into the concrete and thereby the life length models for concrete structures....

  15. DESCRIPTION OF THE TRITIUM-PRODUCING BURNABLE ABSORBER ROD FOR THE COMMERCIAL LIGHT WATER REACTOR TTQP-1-015 Rev 19

    Energy Technology Data Exchange (ETDEWEB)

    Burns, Kimberly A.; Love, Edward F.; Thornhill, Cheryl K.

    2012-02-01

    Tritium-producing burnable absorber rods (TPBARs) used in the U.S. Department of Energy’s Tritium Readiness Program are designed to produce tritium when placed in a Westinghouse or Framatome 17x17 fuel assembly and irradiated in a pressurized water reactor (PWR). This document provides an unclassified description of the current design baseline for the TPBARs. This design baseline is currently valid only for Watts Bar reactor production cores. A description of the Lead Use TPBARs will not be covered in the text of the document, but the applicable drawings, specifications and test plan will be included in the appropriate appendices.

  16. Clean-up system for pool water in pressure suppression chamber and operation method therefor

    Energy Technology Data Exchange (ETDEWEB)

    Hirabayashi, Kentaro; Kinoshita, Shoichiro

    1996-09-17

    Pool water in a pressure suppression chamber of a BWR type reactor is sucked by a pump of an after-heat removing system. The pool water pressurized here is sent to the pressure suppression chamber by way of a heat exchanger and a test line backwarding pipeline to stir the pool water in the pressure suppression chamber. Further, the pool water pressurized by the pump is sent to the pressure suppression chamber by way of a filtration desalting device and an exit pipe to purify the pool water. Upon cleaning of pipelines before the start of a periodical test, the pool water sucked by the pump is sent to the filtration desalting device and recovered to the pressure suppression chamber. This can reduce the amount of impurities carried to the suppression chamber. After the cleaning of the pipelines, pool water is passed through the test line backwarding pipeline, so that the pool water can be stirred at the same time. (I.N.)

  17. Pressure probe study of the water relations of Phycomyces blakesleeanus sporangiophores

    Science.gov (United States)

    Cosgrove, D. J.; Ortega, J. K.; Shropshire, W. Jr

    1987-01-01

    The physical characteristics which govern the water relations of the giant-celled sporangiophore of Phycomyces blakesleeanus were measured with the pressure probe technique and with nanoliter osmometry. These properties are important because they govern water uptake associated with cell growth and because they may influence expansion of the sporangiophore wall. Turgor pressure ranged from 1.1 to 6.6 bars (mean = 4.1 bars), and was the same for stage I and stage IV sporangiophores. Sporangiophore osmotic pressure averaged 11.5 bars. From the difference between cell osmotic pressure and turgor pressure, the average water potential of the sporangiophore was calculated to be about -7.4 bars. When sporangiophores were submerged under water, turgor remained nearly constant. We propose that the low cell turgor pressure is due to solutes in the cell wall solution, i.e., between the cuticle and the plasma membrane. Membrane hydraulic conductivity averaged 4.6 x 10(-6) cm s-1 bar-1, and was significantly greater in stage I sporangiophores than in stage IV sporangiophores. Contrary to previous reports, the sporangiophore is separated from the supporting mycelium by septa which prevent bulk volume flow between the two regions. The presence of a wall compartment between the cuticle and the plasma membrane results in anomalous osmosis during pressure clamp measurements. This behavior arises because of changes in solute concentration as water moves into or out of the wall compartment surrounding the sporangiophore. Theoretical analysis shows how the equations governing transient water flow are altered by the characteristics of the cell wall compartment.

  18. Experiments on aerosol removal by high-pressure water spray

    Energy Technology Data Exchange (ETDEWEB)

    Corno, Ada del, E-mail: delcorno@rse-web.it [RSE, Power Generation Technologies and Materials Dept, via Rubattino 54, I-20134 Milano (Italy); Morandi, Sonia, E-mail: morandi@rse-web.it [RSE, Power Generation Technologies and Materials Dept, via Rubattino 54, I-20134 Milano (Italy); Parozzi, Flavio, E-mail: parozzi@rse-web.it [RSE, Power Generation Technologies and Materials Dept, via Rubattino 54, I-20134 Milano (Italy); Araneo, Lucio, E-mail: lucio.araneo@polimi.it [Politecnico di Milano, Department of Energy, via Lambruschini 4A, I-20156 Milano (Italy); CNR-IENI, via Cozzi 53, I-20125 Milano (Italy); Casella, Francesco, E-mail: francesco2.casella@mail.polimi.it [Politecnico di Milano, Department of Energy, via Lambruschini 4A, I-20156 Milano (Italy)

    2017-01-15

    Highlights: • Experimental research to measure the efficiency of high-pressure sprays in capturing aerosols if applied to a filtered containment venting system in case of severe accident. • Cloud of monodispersed SiO{sub 2} particles with sizes 0.5 or 1.0 μm and initial concentration in the range 2–90 mg/m{sup 3}. • Carried out in a chamber 0.5 × 1.0 m and 1.5 m high, with transparent walls equipped with a high pressure water spray with single nozzle. • Respect to low-pressure sprays, removal efficiency turned out significant: the half-life for 1 μm particles with a removal high-pressure spray system is orders of magnitude shorter than that with a low-pressure sprays system. - Abstract: An experimental research was managed in the framework of the PASSAM European Project to measure the efficiency of high-pressure sprays in capturing aerosols when applied to a filtered containment venting system in case of severe accident. The campaign was carried out in a purposely built facility composed by a scrubbing chamber 0.5 × 1.0 m and 1.5 m high, with transparent walls to permit the complete view of the aerosol removal process, where the aerosol was injected to form a cloud of specific particle concentration. The chamber was equipped with a high pressure water spray system with a single nozzle placed on its top. The test matrix consisted in the combination of water pressure injections, in the range 50–130 bar, on a cloud of monodispersed SiO{sub 2} particles with sizes 0.5 or 1.0 μm and initial concentration ranging between 2 and 99 mg/m{sup 3}. The spray was kept running for 2 min and the efficiency of the removal was evaluated, along the test time, using an optical particle sizer. With respect to low-pressure sprays, the removal efficiency turned out much more significant: the half-life for 1 μm particles with a removal high-pressure spray system is orders of magnitude shorter than that with a low-pressure spray system. The highest removal rate was

  19. Experiments on aerosol removal by high-pressure water spray

    International Nuclear Information System (INIS)

    Corno, Ada del; Morandi, Sonia; Parozzi, Flavio; Araneo, Lucio; Casella, Francesco

    2017-01-01

    Highlights: • Experimental research to measure the efficiency of high-pressure sprays in capturing aerosols if applied to a filtered containment venting system in case of severe accident. • Cloud of monodispersed SiO_2 particles with sizes 0.5 or 1.0 μm and initial concentration in the range 2–90 mg/m"3. • Carried out in a chamber 0.5 × 1.0 m and 1.5 m high, with transparent walls equipped with a high pressure water spray with single nozzle. • Respect to low-pressure sprays, removal efficiency turned out significant: the half-life for 1 μm particles with a removal high-pressure spray system is orders of magnitude shorter than that with a low-pressure sprays system. - Abstract: An experimental research was managed in the framework of the PASSAM European Project to measure the efficiency of high-pressure sprays in capturing aerosols when applied to a filtered containment venting system in case of severe accident. The campaign was carried out in a purposely built facility composed by a scrubbing chamber 0.5 × 1.0 m and 1.5 m high, with transparent walls to permit the complete view of the aerosol removal process, where the aerosol was injected to form a cloud of specific particle concentration. The chamber was equipped with a high pressure water spray system with a single nozzle placed on its top. The test matrix consisted in the combination of water pressure injections, in the range 50–130 bar, on a cloud of monodispersed SiO_2 particles with sizes 0.5 or 1.0 μm and initial concentration ranging between 2 and 99 mg/m"3. The spray was kept running for 2 min and the efficiency of the removal was evaluated, along the test time, using an optical particle sizer. With respect to low-pressure sprays, the removal efficiency turned out much more significant: the half-life for 1 μm particles with a removal high-pressure spray system is orders of magnitude shorter than that with a low-pressure spray system. The highest removal rate was detected with 1

  20. 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

  1. Sloshing of water in annular pressure-suppression pool of boiling water reactors under earthquake ground motions

    International Nuclear Information System (INIS)

    Aslam, M.; Godden, W.G.; Scalise, D.T.

    1979-10-01

    This report presents an analytical investigation of the sloshing response of water in annular-circular as well as simple-circular tanks under horizontal earthquake ground motions, and the results are verified with tests. This study was motivated because of the use of annular tanks for pressure-suppression pools in Boiling Water Reactors. Such a pressure-suppression pool would typically have 80 ft and 120 ft inside and outside diameters and a water depth of 20 ft. The analysis was based upon potential flow theory and a computer program was written to obtain time-history plots of sloshing displacements of water and the dynamic pressures. Tests were carried out on 1/80th and 1/15th scale models under sinusoidal as well as simulated earthquake ground motions. Tests and analytical results regarding the natural frequencies, surface water displacements, and dynamic pressures were compared and a good agreement was found for relatively small displacements. The computer program gave satisfactory results as long as the maximum water surface displacements were less than 30 in., which is roughly the value obtained under full intensity of El Centro earthquake

  2. Improvement in fuel utilization in pressurized heavy water reactors due to increased heavy water purity

    International Nuclear Information System (INIS)

    Balakrishnan, M.R.

    1991-01-01

    This paper reports that in a pressurized heavy water reactor (PHWR), the reactivity of the reactor and, consequently, the discharge burnup of the fuel depend on the isotopic purity of the heavy water used in the reactor. The optimal purity of heavy water used in PHWRs, in turn, depends on the cost of fabricated uranium fuel and on the incremental cost incurred in improving the heavy water purity. The physics and economics aspects of the desirability of increasing the heavy water purity in PHWRs in India were first examined in 1978. With the cost data available at that time, it was found that improving the heavy water purity from 99.80% to 99.95% was economically attractive. The same problem is reinvestigated with current cost data. Even now, there is sufficient incentive to improve the isotopic purity of heavy water used in PHWRs. Admittedly, the economic advantage that can be derived depends on the cost of the fabricated fuel. Nevertheless, irrespective of the economics, there is also a fairly substantial saving in natural uranium. That the increase in the heavy water purity is to be maintained only in the low-pressure moderator system, and not in the high-pressure coolant system, makes the option of achieving higher fuel burnup with higher heavy water purity feasible

  3. Fundamentals of pressurized water reactors

    International Nuclear Information System (INIS)

    Murray, L.

    1982-01-01

    In many countries, the pressurized water reactor (PWR) is the most widely used, even though it requires enrichment of the uranium to about 3% in U-235 and the moderator-coolant must be maintained at a high pressure, about 2200 pounds per square inch. Our objective in this series of seven lectures is to describe the design and operating characteristics of the PWR system, discuss the reactor physics methods used to evaluate performance, examine the way fuel is consumed and produced, study the instrumentation system, review the physics measurements made during initial startup of the reactor, and outline the administrative aspects of starting up a reactor and operating it safely and effectively

  4. 46 CFR 52.01-110 - Water-level indicators, water columns, gauge-glass connections, gauge cocks, and pressure gauges...

    Science.gov (United States)

    2010-10-01

    ... § 52.01-110 Water-level indicators, water columns, gauge-glass connections, gauge cocks, and pressure... 46 Shipping 2 2010-10-01 2010-10-01 false Water-level indicators, water columns, gauge-glass connections, gauge cocks, and pressure gauges (modifies PG-60). 52.01-110 Section 52.01-110 Shipping COAST...

  5. Evaluating the Laplace pressure of water nanodroplets from simulations

    Science.gov (United States)

    Malek, Shahrazad M. A.; Sciortino, Francesco; Poole, Peter H.; Saika-Voivod, Ivan

    2018-04-01

    We calculate the components of the microscopic pressure tensor as a function of radial distance r from the centre of a spherical water droplet, modelled using the TIP4P/2005 potential. To do so, we modify a coarse-graining method for calculating the microscopic pressure (Ikeshoji et al 2003 Mol. Simul. 29 101) in order to apply it to a rigid molecular model of water. As test cases, we study nanodroplets ranging in size from 776 to 2880 molecules at 220 K. Beneath a surface region comprising approximately two molecular layers, the pressure tensor becomes approximately isotropic and constant with r. We find that the dependence of the pressure on droplet radius is that expected from the Young-Laplace equation, despite the small size of the droplets.

  6. 75 FR 77010 - Nextera Energy Point Beach, LLC; Point Beach Nuclear Plant, Units 1 and 2, Draft Environmental...

    Science.gov (United States)

    2010-12-10

    ... waste is low-level radioactive waste which includes sludge, oily waste, bead resin, spent filters, and... Impact of Transportation of Fuel and Waste to and from One Light-Water-Cooled Nuclear Power Reactor..., wetlands, and open areas. Each of the two units at PBNP use Westinghouse pressurized water reactors...

  7. LINEAR KERNEL SUPPORT VECTOR MACHINES FOR MODELING PORE-WATER PRESSURE RESPONSES

    Directory of Open Access Journals (Sweden)

    KHAMARUZAMAN W. YUSOF

    2017-08-01

    Full Text Available Pore-water pressure responses are vital in many aspects of slope management, design and monitoring. Its measurement however, is difficult, expensive and time consuming. Studies on its predictions are lacking. Support vector machines with linear kernel was used here to predict the responses of pore-water pressure to rainfall. Pore-water pressure response data was collected from slope instrumentation program. Support vector machine meta-parameter calibration and model development was carried out using grid search and k-fold cross validation. The mean square error for the model on scaled test data is 0.0015 and the coefficient of determination is 0.9321. Although pore-water pressure response to rainfall is a complex nonlinear process, the use of linear kernel support vector machine can be employed where high accuracy can be sacrificed for computational ease and time.

  8. Microwave measurements of water vapor partial pressure at high temperatures

    International Nuclear Information System (INIS)

    Latorre, V.R.

    1991-01-01

    One of the desired parameters in the Yucca Mountain Project is the capillary pressure of the rock comprising the repository. This parameter is related to the partial pressure of water vapor in the air when in equilibrium with the rock mass. Although there are a number of devices that will measure the relative humidity (directly related to the water vapor partial pressure), they generally will fail at temperatures on the order of 150C. Since thee author has observed borehole temperatures considerably in excess of this value in G-Tunnel at the Nevada Test Site (NTS), a different scheme is required to obtain the desired partial pressure data at higher temperatures. This chapter presents a microwave technique that has been developed to measure water vapor partial pressure in boreholes at temperatures up to 250C. The heart of the system is a microwave coaxial resonator whose resonant frequency is inversely proportional to the square root of the real part of the complex dielectric constant of the medium (air) filling the resonator. The real part of the dielectric constant of air is approximately equal to the square of the refractive index which, in turn, is proportional to the partial pressure of the water vapor in the air. Thus, a microwave resonant cavity can be used to measure changes in the relative humidity or partial pressure of water vapor in the air. Since this type of device is constructed of metal, it is able to withstand very high temperatures. The actual limitation is the temperature limit of the dielectric material in the cable connecting the resonator to its driving and monitoring equipment-an automatic network analyzer in our case. In the following sections, the theory of operation, design, construction, calibration and installation of the microwave diagnostics system is presented. The results and conclusions are also presented, along with suggestions for future work

  9. Frictional pressure drop of high pressure steam-water two-phase flow in internally helical ribbed tubes

    International Nuclear Information System (INIS)

    Tingkuan, C.; Xuanzheng, C.

    1987-01-01

    It is well known that the internally helical ribbed tubes are effective in suppressing the dry-out in boiling tubes at high pressures, so they are widely used as furnace water wall tubes in modern large steam power boilers. Design of the boilers requires the data on frictional pressure drop characteristics of the ribbed tubes, but they are not sufficient now. This paper describes the experimental results on the adiabatic frictional pressure drop in both horizontal ribbed tubes with measured mean inside diameter of 11.69 mm and 35.42 mm at high pressure from 10 to 21 MPa, mass flow rate from 350 to 3800 kg/m/sup 2/s and steam quality from 0 to 1 in our high pressure electrically heated water loop. Simultaneously, both smooth tubes under the same conditions for comparison. Based on the tests the correlation for determining the frictional pressure drop of internally ribbed tubes are proposed

  10. 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)

  11. 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

  12. The future 700 MWe pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Bhardwaj, S.A.

    2006-01-01

    The design of a 700 MWe pressurized heavy water reactor has been developed. The design is based on the twin 540 MWe reactors at Tarapur of which the first unit has been made critical in less than 5 years from construction commencement. In the 700 MWe design boiling of the coolant, to a limited extent, has been allowed near the channel exit. While making the plant layout more compact, emphasis has been on constructability. Saving in capital cost of about 15%, over the present units, is expected. The paper describes salient design features of 700 MWe pressurized heavy water reactor

  13. Assessment of Analytic Water hammer Pressure Model of FAI/08-70

    International Nuclear Information System (INIS)

    Park, Ju Yeop; Yoo, Seung Hun; Seul, Kwang-Won

    2016-01-01

    In evaluating water hammer effect on the safety related systems, methods developed by the US utility are likely to be adopted in Korea. For example, the US utility developed specific methods to evaluate pressure and loading transient on piping due to water hammer as in FAI/08-70. The methods of FAI/08-70 would be applied in Korea when any regulatory request on the evaluation of water hammer effect due to the non-condensable gas accumulation in the safety related systems. Specifically, FAI/08-70 gives an analytic model which can be used to analyze the maximum transient pressure and maximum transient loading on the piping of the safety-related systems due to the non-condensable induced water hammer effect. Therefore, it seems to be meaningful to review the FAI/08-70 methods and attempt to apply the methods to a specific case to see if they really give reasonable estimate before the application of FAI/08-70 methods to domestic nuclear power plants. In the present study, analytic water hammer pressure model of FAI/08-70 is reviewed in detail and the model is applied to the specific experiment of FAI/08-70 to see if the analytic water hammer pressure model really gives reasonable estimate of the peak water hammer pressure. Specifically, we assess the experiment 52A of FAI/08-70 which adopts flushed initial condition with a short rising piping length and a high level piping length of 51inch. The calculated analytic water hammer pressure peak shows a close agreement with the measured experimental data of 52A. Unfortunately, however, the theoretical value is a little bit less than that of the experimental value. This implies the analytic model of FAI/08-70 is not conservative

  14. Assessment of Analytic Water hammer Pressure Model of FAI/08-70

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ju Yeop; Yoo, Seung Hun; Seul, Kwang-Won [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    In evaluating water hammer effect on the safety related systems, methods developed by the US utility are likely to be adopted in Korea. For example, the US utility developed specific methods to evaluate pressure and loading transient on piping due to water hammer as in FAI/08-70. The methods of FAI/08-70 would be applied in Korea when any regulatory request on the evaluation of water hammer effect due to the non-condensable gas accumulation in the safety related systems. Specifically, FAI/08-70 gives an analytic model which can be used to analyze the maximum transient pressure and maximum transient loading on the piping of the safety-related systems due to the non-condensable induced water hammer effect. Therefore, it seems to be meaningful to review the FAI/08-70 methods and attempt to apply the methods to a specific case to see if they really give reasonable estimate before the application of FAI/08-70 methods to domestic nuclear power plants. In the present study, analytic water hammer pressure model of FAI/08-70 is reviewed in detail and the model is applied to the specific experiment of FAI/08-70 to see if the analytic water hammer pressure model really gives reasonable estimate of the peak water hammer pressure. Specifically, we assess the experiment 52A of FAI/08-70 which adopts flushed initial condition with a short rising piping length and a high level piping length of 51inch. The calculated analytic water hammer pressure peak shows a close agreement with the measured experimental data of 52A. Unfortunately, however, the theoretical value is a little bit less than that of the experimental value. This implies the analytic model of FAI/08-70 is not conservative.

  15. Development of Extended Period Pressure-Dependent Demand Water Distribution Models

    Energy Technology Data Exchange (ETDEWEB)

    Judi, David R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mcpherson, Timothy N. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-03-20

    Los Alamos National Laboratory (LANL) has used modeling and simulation of water distribution systems for N-1 contingency analyses to assess criticality of water system assets. Critical components considered in these analyses include pumps, tanks, and supply sources, in addition to critical pipes or aqueducts. A contingency represents the complete removal of the asset from system operation. For each contingency, an extended period simulation (EPS) is run using EPANET. An EPS simulates water system behavior over a time period, typically at least 24 hours. It assesses the ability of a system to respond and recover from asset disruption through distributed storage in tanks throughout the system. Contingencies of concern are identified as those in which some portion of the water system has unmet delivery requirements. A delivery requirement is defined as an aggregation of water demands within a service area, similar to an electric power demand. The metric used to identify areas of unmet delivery requirement in these studies is a pressure threshold of 15 pounds per square inch (psi). This pressure threshold is used because it is below the required pressure for fire protection. Any location in the model with pressure that drops below this threshold at any time during an EPS is considered to have unmet service requirements and is used to determine cascading consequences. The outage area for a contingency is the aggregation of all service areas with a pressure below the threshold at any time during the EPS.

  16. 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)

  17. Ultra-high pressure water jetting for coating removal and surface preparation

    Science.gov (United States)

    Johnson, Spencer T.

    1995-01-01

    This paper shall examine the basics of water technology with particular attention paid to systems currently in use and some select new applications. By providing an overview of commercially available water jet systems in the context of recent case histories, potential users may evaluate the process for future applications. With the on going introduction of regulations prohibiting the use of chemical paint strippers, manual scrapping and dry abrasive media blasting, the need for an environmentally compliant coating removal process has been mandated. Water jet cleaning has been a traditional part of many industrial processed for year, although it has only been in the last few years that reliable pumping equipment capable of ultra-high pressure operation have become available. With the advent of water jet pumping equipment capable of sustaining pressures in excess of 36,000 psi. there has been shift away from lower pressure, high water volume systems. One of the major factors in driving industry to seek higher pressures is the ability to offer higher productivity rates while lowering the quantity of water used and subsequently reprocessed. Among benefits of the trend toward higher pressure/lower volume systems is the corresponding reduction in water jet reaction forces making hand held water jetting practical and safe. Other unique applications made possible by these new generation pumping systems include the use of alternative fluids including liquid ammonia for specialized and hazardous material removal applications. A review of the equipment used and the required modifications will be presented along with the conclusions reached reached during this test program.

  18. Electrochemical noise measurements under pressurized water reactor conditions

    International Nuclear Information System (INIS)

    Van Nieuwenhove, R.

    2000-01-01

    Electrochemical potential noise measurements on sensitized stainless steel pressure tubes under pressurized water reactor (PWR) conditions were performed for the first time. Very short potential spikes, believed to be associated to crack initiation events, were detected when stressing the sample above the yield strength and increased in magnitude until the sample broke. Sudden increases of plastic deformation, as induced by an increased tube pressure, resulted in slower, high-amplitude potential transients, often accompanied by a reduction in noise level

  19. 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

  20. Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Krishnan, S.; Bhasin, V.; Mahajan, S.C.

    1997-01-01

    Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300 degrees C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered

  1. Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Krishnan, S.; Bhasin, V.; Mahajan, S.C. [Bhabha Atomic Research Centre, Bombay (India)] [and others

    1997-04-01

    Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300{degrees}C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered.

  2. Low internal pressure in femtoliter water capillary bridges reduces evaporation rates.

    Science.gov (United States)

    Cho, Kun; Hwang, In Gyu; Kim, Yeseul; Lim, Su Jin; Lim, Jun; Kim, Joon Heon; Gim, Bopil; Weon, Byung Mook

    2016-03-01

    Capillary bridges are usually formed by a small liquid volume in a confined space between two solid surfaces. They can have a lower internal pressure than the surrounding pressure for volumes of the order of femtoliters. Femtoliter capillary bridges with relatively rapid evaporation rates are difficult to explore experimentally. To understand in detail the evaporation of femtoliter capillary bridges, we present a feasible experimental method to directly visualize how water bridges evaporate between a microsphere and a flat substrate in still air using transmission X-ray microscopy. Precise measurements of evaporation rates for water bridges show that lower water pressure than surrounding pressure can significantly decrease evaporation through the suppression of vapor diffusion. This finding provides insight into the evaporation of ultrasmall capillary bridges.

  3. Low internal pressure in femtoliter water capillary bridges reduces evaporation rates

    Science.gov (United States)

    Cho, Kun; Hwang, In Gyu; Kim, Yeseul; Lim, Su Jin; Lim, Jun; Kim, Joon Heon; Gim, Bopil; Weon, Byung Mook

    2016-01-01

    Capillary bridges are usually formed by a small liquid volume in a confined space between two solid surfaces. They can have a lower internal pressure than the surrounding pressure for volumes of the order of femtoliters. Femtoliter capillary bridges with relatively rapid evaporation rates are difficult to explore experimentally. To understand in detail the evaporation of femtoliter capillary bridges, we present a feasible experimental method to directly visualize how water bridges evaporate between a microsphere and a flat substrate in still air using transmission X-ray microscopy. Precise measurements of evaporation rates for water bridges show that lower water pressure than surrounding pressure can significantly decrease evaporation through the suppression of vapor diffusion. This finding provides insight into the evaporation of ultrasmall capillary bridges. PMID:26928329

  4. Effect of Water Cut on Pressure Drop of Oil (D130) -Water Flow in 4″Horizontal Pipe

    Science.gov (United States)

    Basha, Mehaboob; Shaahid, S. M.; Al-Hems, Luai M.

    2018-03-01

    The oil-water flow in pipes is a challenging subject that is rich in physics and practical applications. It is often encountered in many oil and chemical industries. The pressure gradient of two phase flow is still subject of immense research. The present study reports pressure measurements of oil (D130)-water flow in a horizontal 4″ diameter stainless steel pipe at different flow conditions. Experiments were carried out for different water cuts (WC); 0-100%. Inlet oil-water flow rates were varied from 4000 to 8000 barrels-per-day in steps of 2000. It has been found that the frictional pressure drop decreases for WC = 0 - 40 %. With further increase in WC, friction pressure drop increases, this could be due to phase inversion.

  5. Low cost sonoluminescence experiment in pressurized water

    Science.gov (United States)

    Bernal, L.; Insabella, M.; Bilbao, L.

    2012-06-01

    We present a low cost design for demostration and mesurements of light emmision from a sonoluminescence experiment. Using presurized water introduced in an acrylic cylinder and one piezoelectric from an ultrasonic cleaner, we are able to generate cavitacion zones with emission of light. The use of argon to pressurize the water improves the emission an the light can be seen at naked eye in a softlit ambient.

  6. Complex cooling water systems optimization with pressure drop consideration

    CSIR Research Space (South Africa)

    Gololo, KV

    2012-12-01

    Full Text Available Pressure drop consideration has shown to be an essential requirement for the synthesis of a cooling water network where reuse/recycle philosophy is employed. This is due to an increased network pressure drop associated with additional reuse...

  7. Documentation of probabilistic fracture mechanics codes used for reactor pressure vessels subjected to pressurized thermal shock loading: Parts 1 and 2. Final report

    International Nuclear Information System (INIS)

    Balkey, K.; Witt, F.J.; Bishop, B.A.

    1995-06-01

    Significant attention has been focused on the issue of reactor vessel pressurized thermal shock (PTS) for many years. Pressurized thermal shock transient events are characterized by a rapid cooldown at potentially high pressure levels that could lead to a reactor vessel integrity concern for some pressurized water reactors. As a result of regulatory and industry efforts in the early 1980's, a probabilistic risk assessment methodology has been established to address this concern. Probabilistic fracture mechanics analyses are performed as part of this methodology to determine conditional probability of significant flaw extension for given pressurized thermal shock events. While recent industry efforts are underway to benchmark probabilistic fracture mechanics computer codes that are currently used by the nuclear industry, Part I of this report describes the comparison of two independent computer codes used at the time of the development of the original U.S. Nuclear Regulatory Commission (NRC) pressurized thermal shock rule. The work that was originally performed in 1982 and 1983 to compare the U.S. NRC - VISA and Westinghouse (W) - PFM computer codes has been documented and is provided in Part I of this report. Part II of this report describes the results of more recent industry efforts to benchmark PFM computer codes used by the nuclear industry. This study was conducted as part of the USNRC-EPRI Coordinated Research Program for reviewing the technical basis for pressurized thermal shock (PTS) analyses of the reactor pressure vessel. The work focused on the probabilistic fracture mechanics (PFM) analysis codes and methods used to perform the PTS calculations. An in-depth review of the methodologies was performed to verify the accuracy and adequacy of the various different codes. The review was structured around a series of benchmark sample problems to provide a specific context for discussion and examination of the fracture mechanics methodology

  8. Organic Tank Safety Project: development of a method to measure the equilibrium water content of Hanford organic tank wastes and demonstration of method on actual waste

    International Nuclear Information System (INIS)

    Scheele, R.D.; Bredt, P.R.; Sell, R.L.

    1996-09-01

    Some of Hanford's underground waste storage tanks contain Organic- bearing high level wastes that are high priority safety issues because of potentially hazardous chemical reactions of organics with inorganic oxidants in these wastes such as nitrates and nitrites. To ensure continued safe storage of these wastes, Westinghouse Hanford Company has placed affected tanks on the Organic Watch List and manages them under special rules. Because water content has been identified as the most efficient agent for preventing a propagating reaction and is an integral part of the criteria developed to ensure continued safe storage of Hanford's organic-bearing radioactive tank wastes, as part of the Organic Tank Safety Program the Pacific Northwest National Laboratory developed and demonstrated a simple and easily implemented procedure to determine the equilibrium water content of these potentially reactive wastes exposed to the range of water vapor pressures that might be experienced during the wastes' future storage. This work focused on the equilibrium water content and did not investigate the various factors such as at sign ventilation, tank surface area, and waste porosity that control the rate that the waste would come into equilibrium, with either the average Hanford water partial pressure 5.5 torr or other possible water partial pressures

  9. Organic Tank Safety Project: development of a method to measure the equilibrium water content of Hanford organic tank wastes and demonstration of method on actual waste

    Energy Technology Data Exchange (ETDEWEB)

    Scheele, R.D.; Bredt, P.R.; Sell, R.L.

    1996-09-01

    Some of Hanford`s underground waste storage tanks contain Organic- bearing high level wastes that are high priority safety issues because of potentially hazardous chemical reactions of organics with inorganic oxidants in these wastes such as nitrates and nitrites. To ensure continued safe storage of these wastes, Westinghouse Hanford Company has placed affected tanks on the Organic Watch List and manages them under special rules. Because water content has been identified as the most efficient agent for preventing a propagating reaction and is an integral part of the criteria developed to ensure continued safe storage of Hanford`s organic-bearing radioactive tank wastes, as part of the Organic Tank Safety Program the Pacific Northwest National Laboratory developed and demonstrated a simple and easily implemented procedure to determine the equilibrium water content of these potentially reactive wastes exposed to the range of water vapor pressures that might be experienced during the wastes` future storage. This work focused on the equilibrium water content and did not investigate the various factors such as @ ventilation, tank surface area, and waste porosity that control the rate that the waste would come into equilibrium, with either the average Hanford water partial pressure 5.5 torr or other possible water partial pressures.

  10. Light-water reactor pressure vessel surveillance standards

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The master matrix standard describes a series of standard practices, guides, and methods for the prediction of neutron-induced changes in light-water reactor (LWR) pressure vessel steels throughout a pressure vessel's service life. Some of these are existing American Society for Testing and Materials (ASTM) standards, some are ASTM standards that have been modified, and some are newly proposed ASTM standards. The current (1) scope, (2) areas of application, (3) interrelationships, and (4) status and time table of development, improvement, validation, and calibration for a series of 16 ASTM standards are defined. The standard also includes a discussion of LWR pressure vessel surveillance - justification, requirements, and status of work

  11. High pressure, low pressure and hot water heating systems in hospitals. Hochdruck-, Niederdruck- und Warmwasserheizungsanlagen im Krankenhaus

    Energy Technology Data Exchange (ETDEWEB)

    Riedle, K [H. Riedle GmbH, Wiesbaden (Germany)

    1994-07-01

    In hospital nowadays the limitation of the use of steam boilers and their direct supply network to the possible minimum is aimed at when the heating system is exchanged or retrofitted. Independent of the fact whether high pressure or low pressure steam or hot water is used the optimum water treatment should be carried out with a minimum of chemical substances. Here hydroquinone, neutralizing amines, carbohydrazide, sodium sulphite and tannins can be used. The dimensioning of hot water heating circuits is shown with examples. (BWI)

  12. On OH production in water containing atmospheric pressure plasmas

    NARCIS (Netherlands)

    Bruggeman, P.J.; Schram, D.C.

    2010-01-01

    In this paper radical production in atmospheric pressure water containing plasmas is discussed. As OH is often an important radical in these discharges the paper focuses on OH production. Besides nanosecond pulsed coronas and diffusive glow discharges, several other atmospheric pressure plasmas

  13. Failure Mode of the Water-filled Fractures under Hydraulic Pressure in Karst Tunnels

    Directory of Open Access Journals (Sweden)

    Dong Xin

    2017-06-01

    Full Text Available Water-filled fractures continue to grow after the excavation of karst tunnels, and the hydraulic pressure in these fractures changes along with such growth. This paper simplifies the fractures in the surrounding rock as flat ellipses and then identifies the critical hydraulic pressure values required for the occurrence of tensile-shear and compression-shear failures in water-filled fractures in the case of plane stress. The occurrence of tensile-shear fracture requires a larger critical hydraulic pressure than compression-shear failure in the same fracture. This paper examines the effects of fracture strike and lateral pressure coefficient on critical hydraulic pressure, and identifies compression-shear failure as the main failure mode of water-filled fractures. This paper also analyses the hydraulic pressure distribution in fractures with different extensions, and reveals that hydraulic pressure decreases along with the continuous growth of fractures and cannot completely fill a newly formed fracture with water. Fracture growth may be interrupted under the effect of hydraulic tensile shear.

  14. 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)

  15. 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)

  16. 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)

  17. Can the water content of highly compacted bentonite be increased by applying a high water pressure?

    International Nuclear Information System (INIS)

    Pusch, R.; Kasbohm, J.

    2001-10-01

    A great many laboratory investigations have shown that the water uptake in highly compacted MX-80 clay takes place by diffusion at low external pressure. It means that wetting of the clay buffer in the deposition holes of a KBS-3 repository is very slow if the water pressure is low and that complete water saturation can take several tens of years if the initial degree of water saturation of the buffer clay and the ability of the rock to give off water are low. It has therefore been asked whether injection of water can raise the degree of water saturation and if a high water pressure in the nearfield can have the same effect. The present report describes attempts to moisten highly compacted blocks of MX-80 clay with a dry density of 1510 kg/m 3 by injecting water under a pressure of 650 kPa through a perforated injection pipe for 3 and 20 minutes, respectively. The interpretation was made by determining the water content of a number of samples located at different distances from the pipe. An attempt to interpret the pattern of distribution of injected uranium acetate solution showed that the channels into which the solution went became closed in a few minutes and that dispersion in the homogenized clay gave low U-concentrations. The result was that the water content increased from about 9 to about 11-12 % within a distance of about 1 centimeter from the injection pipe and to slightly more than 9 % at a distance of about 4-5 cm almost independently of the injection time. Complete water saturation corresponds to a water content of about 30 % and the wetting effect was hence small from a practical point of view. By use of microstructural models it can be shown that injected water enters only the widest channels that remain after the compaction and that these channels are quickly closed by expansion of the hydrating surrounding clay. Part of the particles that are thereby released become transported by the flowing water and cause clogging of the channels, which is

  18. Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Vito, D.J.

    1980-12-01

    The Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors (CE-STS) is a generic document prepared by the US NRC for use in the licensing process of current Combustion Engineering Pressurized Water Reactors. The CE-STS sets forth the limits, operating conditions, and other requirements applicable to nuclear reactor facility operation as set forth by Section 50.36 of 10 CFR 50 for the protection of the health and safety of the public. The document is revised periodically to reflect current licensing requirements

  19. Water cycle and its management for plant habitats at reduced pressures

    Science.gov (United States)

    Rygalov, Vadim Y.; Fowler, Philip A.; Wheeler, Raymond M.; Bucklin, Ray A.

    2004-01-01

    Experimental and mathematical models were developed for describing and testing temperature and humidity parameters for plant production in bioregenerative life support systems. A factor was included for analyzing systems operating at low (10-101.3 kPa) pressure to reduce gas leakage and structural mass (e.g., inflatable greenhouses for space application). The expected close relationship between temperature and relative humidity was observed, along with the importance of heat exchanger coil temperature and air circulation rate. The presence of plants in closed habitats results in increased water flux through the system. Changes in pressure affect gas diffusion rates and surface boundary layers, and change convective transfer capabilities and water evaporation rates. A consistent observation from studies with plants at reduced pressures is increased evapotranspiration rates, even at constant vapor pressure deficits. This suggests that plant water status is a critical factor for managing low-pressure production systems. The approach suggested should help space mission planners design artificial environments in closed habitats.

  20. Design of virtual SCADA simulation system for pressurized water reactor

    International Nuclear Information System (INIS)

    Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman

    2016-01-01

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor

  1. Design of virtual SCADA simulation system for pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wijaksono, Umar, E-mail: umar.wijaksono@student.upi.edu; Abdullah, Ade Gafar; Hakim, Dadang Lukman [Electrical Power System Research Group, Department of Electrical Engineering Education, Jl. Dr. Setiabudi No. 207 Bandung, Indonesia 40154 (Indonesia)

    2016-02-08

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  2. The Oxidation Rate of SiC in High Pressure Water Vapor Environments

    Science.gov (United States)

    Opila, Elizabeth J.; Robinson, R. Craig

    1999-01-01

    CVD SiC and sintered alpha-SiC samples were exposed at 1316 C in a high pressure burner rig at total pressures of 5.7, 15, and 25 atm for times up to 100h. Variations in sample emittance for the first nine hours of exposure were used to determine the thickness of the silica scale as a function of time. After accounting for volatility of silica in water vapor, the parabolic rate constants for Sic in water vapor pressures of 0.7, 1.8 and 3.1 atm were determined. The dependence of the parabolic rate constant on the water vapor pressure yielded a power law exponent of one. Silica growth on Sic is therefore limited by transport of molecular water vapor through the silica scale.

  3. Evaluation of the nucledyne passive containment system

    International Nuclear Information System (INIS)

    1981-04-01

    This reports contains: (1) an evaluation by Gilbert/Commonwealth (G/C) of the NucleDyne passive Containment System (PCS) as that conceptual design is applied to a Westinghouse, two loop, Pressurized Water Reactor; (2) an evaluation by Westinghouse of two questions about the impact of the PCS on the Nuclear Steam Supply System (NSSS), which were posed by G/C and best answered by an NSSS vendor; and (3) replies to both the Gilbert/Commonwealth report and the Westinghoue report by NucleDyne Engineering Corporation

  4. 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

  5. Comparison of air-charged and water-filled urodynamic pressure measurement catheters.

    Science.gov (United States)

    Cooper, M A; Fletter, P C; Zaszczurynski, P J; Damaser, M S

    2011-03-01

    Catheter systems are utilized to measure pressure for diagnosis of voiding dysfunction. In a clinical setting, patient movement and urodynamic pumps introduce hydrostatic and motion artifacts into measurements. Therefore, complete characterization of a catheter system includes its response to artifacts as well its frequency response. The objective of this study was to compare the response of two disposable clinical catheter systems: water-filled and air-charged, to controlled pressure signals to assess their similarities and differences in pressure transduction. We characterized frequency response using a transient step test, which exposed the catheters to a sudden change in pressure; and a sinusoidal frequency sweep test, which exposed the catheters to a sinusoidal pressure wave from 1 to 30 Hz. The response of the catheters to motion artifacts was tested using a vortex and the response to hydrostatic pressure changes was tested by moving the catheter tips to calibrated heights. Water-filled catheters acted as an underdamped system, resonating at 10.13 ± 1.03 Hz and attenuating signals at frequencies higher than 19 Hz. They demonstrated significant motion and hydrostatic artifacts. Air-charged catheters acted as an overdamped system and attenuated signals at frequencies higher than 3.02 ± 0.13 Hz. They demonstrated significantly less motion and hydrostatic artifacts than water-filled catheters. The transient step and frequency sweep tests gave comparable results. Air-charged and water-filled catheters respond to pressure changes in dramatically different ways. Knowledge of the characteristics of the pressure-measuring system is essential to finding the best match for a specific application. Copyright © 2011 Wiley-Liss, Inc.

  6. Basal interstitial water pressure in laboratory debris flows over a rigid bed in an open channel

    Directory of Open Access Journals (Sweden)

    N. Hotta

    2012-08-01

    Full Text Available Measuring the interstitial water pressure of debris flows under various conditions gives essential information on the flow stress structure. This study measured the basal interstitial water pressure during debris flow routing experiments in a laboratory flume. Because a sensitive pressure gauge is required to measure the interstitial water pressure in shallow laboratory debris flows, a differential gas pressure gauge with an attached diaphragm was used. Although this system required calibration before and after each experiment, it showed a linear behavior and a sufficiently high temporal resolution for measuring the interstitial water pressure of debris flows. The values of the interstitial water pressure were low. However, an excess of pressure beyond the hydrostatic pressure was observed with increasing sediment particle size. The measured excess pressure corresponded to the theoretical excess interstitial water pressure, derived as a Reynolds stress in the interstitial water of boulder debris flows. Turbulence was thought to induce a strong shear in the interstitial space of sediment particles. The interstitial water pressure in boulder debris flows should be affected by the fine sediment concentration and the phase transition from laminar to turbulent debris flow; this should be the subject of future studies.

  7. Methodology for surge pressure evaluation in a water injection system

    Energy Technology Data Exchange (ETDEWEB)

    Meliande, Patricia; Nascimento, Elson A. [Universidade Federal Fluminense (UFF), Niteroi, RJ (Brazil). Dept. de Engenharia Civil; Mascarenhas, Flavio C.B. [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Lab. de Hidraulica Computacional; Dandoulakis, Joao P. [SHELL of Brazil, Rio de Janeiro, RJ (Brazil)

    2009-07-01

    Predicting transient effects, known as surge pressures, is of high importance for offshore industry. It involves detailed computer modeling that attempts to simulate the complex interaction between flow line and fluid in order to ensure efficient system integrity. Platform process operators normally raise concerns whether the water injection system is adequately designed or not to be protected against possible surge pressures during sudden valve closure. This report aims to evaluate the surge pressures in Bijupira and Salema water injection systems due to valve closure, through a computer model simulation. Comparisons among the results from empirical formulations are discussed and supplementary analysis for Salema system were performed in order to define the maximum volumetric flow rate for which the design pressure was able to withstand. Maximum surge pressure values of 287.76 bar and 318.58 bar, obtained in Salema and Bijupira respectively, using empirical formulations have surpassed the operating pressure design, while the computer model results have pointed the greatest surge pressure value of 282 bar in Salema system. (author)

  8. Vegetative Propagule Pressure and Water Depth Affect Biomass and Evenness of Submerged Macrophyte Communities.

    Science.gov (United States)

    Li, Hong-Li; Wang, Yong-Yang; Zhang, Qian; Wang, Pu; Zhang, Ming-Xiang; Yu, Fei-Hai

    2015-01-01

    Vegetative propagule pressure may affect the establishment and structure of aquatic plant communities that are commonly dominated by plants capable of clonal growth. We experimentally constructed aquatic communities consisting of four submerged macrophytes (Hydrilla verticillata, Ceratophyllum demersum, Elodea nuttallii and Myriophyllum spicatum) with three levels of vegetative propagule pressure (4, 8 and 16 shoot fragments for communities in each pot) and two levels of water depth (30 cm and 70 cm). Increasing vegetative propagule pressure and decreasing water level significantly increased the growth of the submerged macrophyte communities, suggesting that propagule pressure and water depth should be considered when utilizing vegetative propagules to re-establish submerged macrophyte communities in degraded aquatic ecosystems. However, increasing vegetative propagule pressure and decreasing water level significantly decreased evenness of the submerged macrophyte communities because they markedly increased the dominance of H. verticillata and E. nuttallii, but had little impact on that of C. demersum and M. spicatum. Thus, effects of vegetative propagule pressure and water depth are species-specific and increasing vegetative propagule pressure under lower water level can facilitate the establishment success of submerged macrophyte communities.

  9. Sloshing of water in torus pressure-suppression pool of boiling water reactors under earthquake ground motions

    International Nuclear Information System (INIS)

    Aslam, M.; Godden, W.G.; Scalise, D.T.

    1978-08-01

    This report presents an analytical and experimental investigation into the sloshing of water in torus tanks under horizontal earthquake ground motions. This study was motivated because of the use of torus tanks for pressure-suppression pools in Boiling Water Reactors. Such a pressure-suppression pool would typically have 80 ft and 140 ft inside and outside diameters, a 30 ft diameter section, and a water depth of 15 ft. A general finite element analysis was developed for all axisymmetric tanks and a computer program was written to obtain time-history plots of sloshing displacements of water and dynamic pressures. Tests were carried out on a 1/60th scale model under sinusoidal as well as simulated earthquake ground motions. Tests and analytical results regarding natural frequencies, surface water displacements, and dynamic pressures were compared and a good agreement was found within the range of displacements studied. The computer program gave satisfactory results within a maximum range of sloshing displacements in the full-size prototype of 30 in. which is greater than the value obtained under the full intensity of the El Centro earthquake (N-S component 1940). The range of linear behavior was studied experimentally by subjecting the torus model to increasing intensities of the El Centro earthquake

  10. High converter pressurized water reactor with heavy water as a coolant

    International Nuclear Information System (INIS)

    Ronen, Y.; Reyev, D.

    1983-01-01

    There is an increasing interest in water breeder and high converter reactors. The increase in the conversion ratio of these reactors is obtained by hardening the neutron spectrum achieved by tightening the reactor's lattice. Another way of hardening the neutron spectrum is to replace the light water with heavy water. Two pressurized water reactor fuel cycles that use heavy water as a coolant are considered. The first fuel cycle is based on plutonium and depleted uranium, and the second cycle is based on plutonium and enriched uranium. The uranium ore and separative work unit (SWU) requirements are calculated as well as the fuel cycle cost. The savings in uranium ore are about40 and 60% and about40% in SWU for both fuel cycles considered

  11. A powerful methodology for reactor vessel pressurized thermal shock analysis

    International Nuclear Information System (INIS)

    Boucau, J.; Mager, T.

    1994-01-01

    The recent operating experience of the Pressurized Water Reactor (PWR) Industry has focused increasing attention on the issue of reactor vessel pressurized thermal shock (PTS). More specifically, the review of the old WWER-type of reactors (WWER 440/230) has indicated a sensitive behaviour to neutron embrittlement. This led already to some remedial actions including safety injection water preheating or vessel annealing. Such measures are usually taken based on the analysis of a selected number of conservative PTS events. Consideration of all postulated cooldown events would draw attention to the impact of operator action and control system effects on reactor vessel PTS. Westinghouse has developed a methodology which couples event sequence analysis with probabilistic fracture mechanics analyses, to identify those events that are of primary concern for reactor vessel integrity. Operating experience is utilized to aid in defining the appropriate event sequences and event frequencies of occurrence for the evaluation. Once the event sequences of concern are identified, detailed deterministic thermal-hydraulic and structural evaluations can be performed to determine the conditions required to minimize the extension of postulated flaws or enhance flaw arrest in the reactor vessel. The results of these analyses can then be used to better define further modifications in vessel and plant system design and to operating procedures. The purpose of the present paper will be to describe this methodology and to show its benefits for decision making. (author). 1 ref., 3 figs

  12. Steam generator for a pressurized-water coolant nuclear reactor

    International Nuclear Information System (INIS)

    Schroeder, H.J.; Berger, W.

    1975-01-01

    A description is given of a steam generator which has a vertical cylindrical housing having a steam output outlet, a horizontal tube sheet closing the lower end of this housing, and an inverted U-shaped tube bundle inside of the housing and having vertical inlet and outlet legs with their ends mounted in the tube sheet. Beneath the tube sheet there are inlet and outlet manifolds for the respective ends of the tube bundle so that pressurized-water coolant from a pressurized-water coolant nuclear reactor can be circulated through the tube bundle

  13. Response of steam-water mixtures to pressure transients

    International Nuclear Information System (INIS)

    Hull, L.M.

    1985-01-01

    During the transition phase of a hypothetical core-disruptive accident in a liquid-metal fast breeder reactor, melting fuel-steel mixtures may begin to boil, resulting in a two-phase mixture of molten reactor fuel and steel vapor. Dispersal of this mixture by pressure transients may prevent recriticality of the fuel material. This paper describes the results of a series of experiments that investigated the response of two-phase mixtures to pressure transients. Simulant fluids (steam/water) were used in a transparent 10.2-cm-dia, 63.5-cm-long acrylic tube. The pressure transient was provided by releasing pressurized nitrogen from a supply tank. The data obtained are in the form of pressure-time records and high-speed movies. The varied parameters are initial void fraction (10% and 40%) and transient pressure magnitude (3.45 and 310 kPa)

  14. "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…

  15. Experimental Study on Peak Pressure of Shock Waves in Quasi-Shallow Water

    Directory of Open Access Journals (Sweden)

    Zhenxiong Wang

    2015-01-01

    Full Text Available Based on the similarity laws of the explosion, this research develops similarity requirements of the small-scale experiments of underwater explosions and establishes a regression model for peak pressure of underwater shock waves under experimental condition. Small-scale experiments are carried out with two types of media at the bottom of the water and for different water depths. The peak pressure of underwater shock waves at different measuring points is acquired. A formula consistent with the similarity law of explosions is obtained and an analysis of the regression precision of the formula confirms its accuracy. Significance experiment indicates that the influence of distance between measuring points and charge on peak pressure of underwater shock wave is the greatest and that of water depth is the least within the range of geometric parameters. An analysis of data from experiments with different media at the bottom of the water reveals an influence on the peak pressure, as the peak pressure of a shock wave in a body of water with a bottom soft mud and rocks is about 1.33 times that of the case where the bottom material is only soft mud.

  16. Automated corrosion fatigue crack growth testing in pressurized water environments

    International Nuclear Information System (INIS)

    Ceschini, L.J.; Liaw, P.K.; Rudd, G.E.; Logsdon, W.A.

    1984-01-01

    This paper describes in detail a novel approach to construct a test facility for developing corrosion fatigue crack growth rate (FCGR) properties in aggressive environments. The environment studied is that of a pressurized water reactor (PWR) at 288 0 C (550 0 F) and 13.8 MPa (200 psig). To expedite data generation, each chamber was designed to accommodate two test specimens. A common water recirculation and pressurization system was employed to service two test chambers. Thus, four fatigue crack propagation rate tests could be conducted simultaneously in the pressurized water environment. The data analysis was automated to minimize the typically high labor costs associated with corrosion fatigue crack propagation testing. Verification FCGR tests conducted on an ASTM A469 rotor steel in a room temperature air environment as well as actual PWR environment FCGR tests performed on an ASTM A533 Grade B Class 2 pressure vessel steel demonstrated that the dual specimen test facility is an excellent system for developing the FCGR properties of materials in adverse environments

  17. Correction of Pressure Drop in Steam and Water System in Performance Test of Boiler

    Science.gov (United States)

    Liu, Jinglong; Zhao, Xianqiao; Hou, Fanjun; Wu, Xiaowu; Wang, Feng; Hu, Zhihong; Yang, Xinsen

    2018-01-01

    Steam and water pressure drop is one of the most important characteristics in the boiler performance test. As the measuring points are not in the guaranteed position and the test condition fluctuation exsits, the pressure drop test of steam and water system has the deviation of measuring point position and the deviation of test running parameter. In order to get accurate pressure drop of steam and water system, the corresponding correction should be carried out. This paper introduces the correction method of steam and water pressure drop in boiler performance test.

  18. Corrosion fatigue cracking behavior of Inconel 690 (TT) in secondary water of pressurized water reactors

    International Nuclear Information System (INIS)

    Xiao Jun; Chen Luyao; Qiu Shaoyu; Chen Yong; Lin Zhenxia; Fu Zhenghong

    2015-01-01

    Inconel 690 (TT) is one of the key materials for tubes of steam generators for pressurized water reactors, where it is susceptible to corrosion fatigue cracking. In this paper, the corrosion fatigue cracking behavior of Inconel 690 (TT) was investigated under small scale yielding conditions, in the simulated secondary water of pressurized water reactor. It was observed that the fatigue crack growth rate was accelerated by a maximum factor up to 3 in the simulated secondary water, comparing to that in room temperature air. In addition, it was found that the accelerating effect was influenced by out-of-plane cracking of corrosion fatigue cracks and also correlated with stress intensity factor range, maximum stress intensity factor and stress ratio. (authors)

  19. Pressure Transient Model of Water-Hydraulic Pipelines with Cavitation

    Directory of Open Access Journals (Sweden)

    Dan Jiang

    2018-03-01

    Full Text Available Transient pressure investigation of water-hydraulic pipelines is a challenge in the fluid transmission field, since the flow continuity equation and momentum equation are partial differential, and the vaporous cavitation has high dynamics; the frictional force caused by fluid viscosity is especially uncertain. In this study, due to the different transient pressure dynamics in upstream and downstream pipelines, the finite difference method (FDM is adopted to handle pressure transients with and without cavitation, as well as steady friction and frequency-dependent unsteady friction. Different from the traditional method of characteristics (MOC, the FDM is advantageous in terms of the simple and convenient computation. Furthermore, the mechanism of cavitation growth and collapse are captured both upstream and downstream of the water-hydraulic pipeline, i.e., the cavitation start time, the end time, the duration, the maximum volume, and the corresponding time points. By referring to the experimental results of two previous works, the comparative simulation results of two computation methods are verified in experimental water-hydraulic pipelines, which indicates that the finite difference method shows better data consistency than the MOC.

  20. Vegetative Propagule Pressure and Water Depth Affect Biomass and Evenness of Submerged Macrophyte Communities.

    Directory of Open Access Journals (Sweden)

    Hong-Li Li

    Full Text Available Vegetative propagule pressure may affect the establishment and structure of aquatic plant communities that are commonly dominated by plants capable of clonal growth. We experimentally constructed aquatic communities consisting of four submerged macrophytes (Hydrilla verticillata, Ceratophyllum demersum, Elodea nuttallii and Myriophyllum spicatum with three levels of vegetative propagule pressure (4, 8 and 16 shoot fragments for communities in each pot and two levels of water depth (30 cm and 70 cm. Increasing vegetative propagule pressure and decreasing water level significantly increased the growth of the submerged macrophyte communities, suggesting that propagule pressure and water depth should be considered when utilizing vegetative propagules to re-establish submerged macrophyte communities in degraded aquatic ecosystems. However, increasing vegetative propagule pressure and decreasing water level significantly decreased evenness of the submerged macrophyte communities because they markedly increased the dominance of H. verticillata and E. nuttallii, but had little impact on that of C. demersum and M. spicatum. Thus, effects of vegetative propagule pressure and water depth are species-specific and increasing vegetative propagule pressure under lower water level can facilitate the establishment success of submerged macrophyte communities.

  1. Application-specific integrated circuit design for a typical pressurized water reactor pressure channel trip

    International Nuclear Information System (INIS)

    Battle, R.E.; Manges, W.W.; Emery, M.S.; Vendermolen, R.I.; Bhatt, S.

    1994-01-01

    This article discusses the use of application-specific integrated circuits (ASICs) in nuclear plant safety systems. ASICs have certain advantages over software-based systems because they can be simple enough to be thoroughly tested, and they can be tailored to replace existing equipment. An architecture to replace a pressurized water reactor pressure channel trip is presented. Methods of implementing digital algorithms are also discussed

  2. Investigation of temperature fluctuation phenomena in a stratified steam-water two-phase flow in a simulating pressurizer spray pipe of a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Miyoshi, Koji, E-mail: miyoshi.koj@inss.co.jp; Takenaka, Nobuyuki; Ishida, Taisuke; Sugimoto, Katsumi

    2017-05-15

    Highlights: • Thermal hydraulics phenomena were discussed in a spray pipe of pressurizer. • Temperature fluctuation was investigated in a stratified steam-water two-phase. • Remarkable liquid temperature fluctuations were observed in the liquid layer. • The observed temperature fluctuations were caused by the internal gravity wave. • The temperature fluctuations decreased with increasing dissolved oxygen. - Abstract: Temperature fluctuation phenomena in a stratified steam-water two-phase flow in a horizontal rectangular duct, which simulate a pressurizer spray pipe of a pressurized water reactor, were studied experimentally. Vertical distributions of the temperature and the liquid velocity were measured with water of various dissolved oxygen concentrations. Large liquid temperature fluctuations were observed when the water was deaerated well and dissolved oxygen concentration was around 10 ppb. The large temperature fluctuations were not observed when the oxygen concentration was higher. It was shown that the observed temperature fluctuations were caused by the internal gravity wave since the Richardson numbers were larger than 0.25 and the temperature fluctuation frequencies were around the Brunt-Väisälä frequencies in the present experimental conditions. The temperature fluctuations decreased by the non-condensable gas since the non-condensable gas suppressed the condensation and the temperature difference in the liquid layer was small.

  3. Pressure of drinking water network on the Meyrin site to be boosted

    CERN Multimedia

    2004-01-01

    In the framework of the refurbishment of CERN's drinking water supply system, the final part of the network on the Meyrin site is to be connected to the pumping station operated by Services Industriels de Genève, bringing about a significant increase in the network pressure of up to 5 bar. This means that from January 2005 onwards, the water pressure in buildings will be increased from 2 - 4 bar to 7 - 9 bar. The TS Department will be checking and upgrading the drinking water supply equipment in toilets and washrooms. All users with devices connected to the water supply system are kindly requested to check that these are compatible with the new pressure levels. More information on the buildings affected, the new pressure levels and the dates on which the changes will come into effect can be found at: https://edms.cern.ch/document/525717/1 Should any equipment under your responsibility be incompatible with the future pressure levels, please contact the Technical Control Room on 72201.

  4. Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2011-07-01

    This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

  5. Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

    International Nuclear Information System (INIS)

    Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R.

    2011-01-01

    This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

  6. Pressure control for minimizing leakage in water distribution systems

    OpenAIRE

    Nourhan Samir; Rawya Kansoh; Walid Elbarki; Amr Fleifle

    2017-01-01

    In the last decades water resources availability has been a major issue on the international agenda. In a situation of worsening scarcity of water resources and the rapidly increasing of water demands, the state of water losses management is part of manâs survival on earth. Leakage in water supply networks makes up a significant amount, sometimes more than 70% of the total water losses. The best practices suggest that pressure management is one of the most effective way to reduce the amount o...

  7. 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

  8. 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.

  9. Comparison between intraocular pressure spikes with water loading and postural change.

    Science.gov (United States)

    Chong, Calum Wk; Wang, Sarah B; Jain, Neeranjali S; Bank, Cassandra S; Singh, Ravjit; Bank, Allan; Francis, Ian C; Agar, Ashish

    2016-12-01

    To compare the agreement between peak intraocular pressures measured through the water drinking test and the supine test, in patients with primary open angle glaucoma. Consecutive, prospective, blinded. Twenty-one patients from the Glaucoma Unit, Prince of Wales Hospital, Sydney, Australia. For the supine test, intraocular pressure was recorded immediately after the patient had lain down and at 20 and 40 min. At the second evaluation, intraocular pressure was measured in each patient after drinking 10 mL/kg body weight of water for the water drinking test. Again, all patients had their intraocular pressure measured at 20 and 40 min (t = 20 and t = 40, respectively). Patients were excluded from the study if they had pre-existing cardiac, renal or pulmonary complications or had concurrent ocular disease or an anatomical abnormality (including angle recession, peripheral anterior synechiae and developmental anomalies of the angle) that may have influenced intraocular pressure. Bland-Altman analysis. Bland-Altman analysis indicated an overall excellent agreement in terms of mean difference between methods (1.0, -1.0 and -0.90 mmHg, at 0, 20 and 40 min, respectively). Further, with the exception of t = 40, all measured time points had 95% confidence intervals within 6.5 mmHg of their mean difference on the Bland-Altman plot. There was close agreement between the intraocular pressure values of the supine test and water drinking test. However, as the water drinking test may be uncomfortable and potentially hazardous, there is potential that the supine test may be a safer and more comfortable alternative. © 2016 Royal Australian and New Zealand College of Ophthalmologists.

  10. 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

  11. Effects of water compressibility on the pressure fluctuation prediction in pump turbine

    International Nuclear Information System (INIS)

    Yin, J L; Wang, D Z; Wang, L Q; Wu, Y L; Wei, X Z

    2012-01-01

    The compressible effect of water is a key factor in transient flows. However, it is always neglected in the unsteady simulations for hydraulic machinery. In light of this, the governing equation of the flow is deduced to combine the compressibility of water, and then simulations with compressible and incompressible considerations to the typical unsteady flow phenomenon (Rotor stator interaction) in a pump turbine model are carried out and compared with each other. The results show that water compressibility has great effects on the magnitude and frequency of pressure fluctuation. As the operating condition concerned, the compressibility of water will induce larger pressure fluctuation, which agrees better with measured data. Moreover, the lower frequency component of the pressure signal can only be captured with the combination of water compressibility. It can be concluded that water compressibility is a fatal factor, which cannot be neglected in the unsteady simulations for pump turbines.

  12. China's coal-fired power plants impose pressure on water resources

    NARCIS (Netherlands)

    Zhang, Xinxin; Liu, Junguo; Tang, Yu; Zhao, Xu; Yang, Hong; Gerbens-Leenes, P.W.; Vliet, van Michelle T.H.; Yan, Jinyue

    2017-01-01

    Coal is the dominant fuel for electricity generation around the world. This type of electricity generation uses large amounts of water, increasing pressure on water resources. This calls for an in-depth investigation in the water-energy nexus of coal-fired electricity generation. In China,

  13. Increase in gas output by active modification of the water pressure regime

    Energy Technology Data Exchange (ETDEWEB)

    Zakirov, S N; Gordon, V Y; Kondrat, R M; Kravtsov, N A; Somov, B Y

    1981-01-01

    Based on gas-hydrodynamic calculations made on a planar model formation, two variants of formation working are examined. In the first variant, the modern ideology of working gas fields with a water pressure regime are simulated. In the second variant, working of the formation is modeled according to the suggested ideology of active modification of the water-pressure regime by operating the flooded gas wells. The calculations made indicate the efficiency of active modification of the water pressure regime from the viewpoint of controlling the fund of E wells, and most important, maximizing the final coefficient of gas bed output.

  14. Pore water pressure response to small and large openings in argillaceous rocks

    International Nuclear Information System (INIS)

    Garitte, B.; Gens, A.; Vaunat, J.; Armand, G.; Conil, N.

    2012-01-01

    Document available in extended abstract form only. In the last decade an important amount of piezometers have been installed in the Bure Underground Rock Laboratory (URL) in the vicinity of ongoing works involving gallery excavations and drilling of boreholes and alveoles both in the major and minor stress directions. Relatively far field piezometers (placed one to four diameters from the excavation wall) showed a qualitatively consistent response at different scales. Here, we investigate whether the pore water pressure response around openings of different scales may be up-scaled. An attempt is made to find a common set of parameters that explains quantitatively the rock response at the different scales. The mechanisms underlying the pore water pressure response around an underground opening are twofold. The first class of mechanisms is usually associated with nearly undrained behaviour and the related pore water pressure changes are induced by the stress redistribution triggered by the creation of the tunnel opening causing a reorientation of the principal stresses and influenced by the initial stress anisotropy. These pore water pressure changes are closely linked to the mechanical constitutive law of the rock and to the damage zone around the opening. The second class of mechanisms is related to the drainage of excess pore water pressure relative to a state governed by the atmospheric water pressure condition prescribed at gallery wall and the water flow law, usually Darcy's. Strong anisotropy effects on the hydraulic response of Callovo-Oxfordian Clay can be observed with reference to Figure 1 that shows the pore pressure response to the drilling of a 150 mm-diameter borehole performed to install a heater for the TER thermal experiment. The borehole is aligned with the major horizontal principal stress. Therefore, in principle, the stress state should be approximately isotropic in a cross section of the borehole. As a matter of fact, however, a degree of

  15. BWR [boiling-water reactor] and PWR [pressurized-water reactor] off-normal event descriptions

    International Nuclear Information System (INIS)

    1987-11-01

    This document chronicles a total of 87 reactor event descriptions for use by operator licensing examiners in the construction of simulator scenarios. Events are organized into four categories: (1) boiling-water reactor abnormal events; (2) boiling-water reactor emergency events; (3) pressurized-water reactor abnormal events; and (4) pressurized-water reactor emergency events. Each event described includes a cover sheet and a progression of operator actions flow chart. The cover sheet contains the following general information: initial plant state, sequence initiator, important plant parameters, major plant systems affected, tolerance ranges, final plant state, and competencies tested. The progression of operator actions flow chart depicts, in a flow chart manner, the representative sequence(s) of expected immediate and subsequent candidate actions, including communications, that can be observed during the event. These descriptions are intended to provide examiners with a reliable, performance-based source of information from which to design simulator scenarios that will provide a valid test of the candidates' ability to safely and competently perform all licensed duties and responsibilities

  16. Functional systems of a pressurized water reactor

    International Nuclear Information System (INIS)

    Heinzel, V.

    1982-01-01

    The main topics, discussed in the present paper, are: - Principle design of the reactor coolant system - reactor pressure vessel with internals - containment design - residual heat removal and emergency cooling systems - nuclear component cooling systems - emergency feed water systems - plant electric power supply system. (orig./RW)

  17. Robustness of parameter-less remote real-time pressure control in water distribution systems

    CSIR Research Space (South Africa)

    Page, Philip R

    2017-06-01

    Full Text Available One way of reducing water leakage, pipe bursts and water consumption in a water distribution system (WDS) is to manage the pressure to be as low as possible. This can be done by adjusting a pressure control valve (PCV) in real-time in order to keep...

  18. Device for automatically operating cooling mode of water in a pressure suppression chamber

    International Nuclear Information System (INIS)

    Sato, Hideyuki.

    1975-01-01

    Object: To provide a system for removing residual heat in a reactor safety system, which can automatically cool water in a pressure suppression chamber when a load on a generator is cut off, so as not to scram the reactor. Structure: When a load cut-off signal is generated by means of rapid closure of a turbine regulating valve or due to the load unbalance relay of generator output, or the like, a sea water pump is started to fully open an outlet valve for the sea water pump, a heat exchanging inlet valve and a minimum crow valve and to fully close a heat exchanging bypass valve. In this manner, cooling water for the heat exchanger is secured to start the pump in the system for removing residual heat, and when the pump discharge pressure is in normal condition, the inlet valve in pressure suppression chamber and the spray valve in the pressure suppression chamber are fully opened to automatically cool water in the pressure suppression chamber. (Hanada, M.)

  19. 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

  20. Variations of free gas content in water during pressure fluctuations

    International Nuclear Information System (INIS)

    Keller, A.; Zielke, W.

    1977-01-01

    In this paper an experimental programme is described in order to determine the influence of the cavitation nuclei distribution on cavitation inception. This programme has been used to measure air bubbles dimensions and number and particularly to determine the influence of quick pressure variations on the size on the number of bubbles in a pipe. An optical device counting scattered light is used as a measuring technique. Gas bubbles go through an optical control volume where they receive a high intensity light beam and scatter the light, then led to a photomultiplier; the signals are sorted and counted according to their size. If the number of nuclei, the dimensions of the control volume and the velocity of the water are known, it is possible to determine bubbles concentrations and the bulk modulus of the water. This measuring technique has been applied to a flow in a 140 mm diameter pipe with quick pressure variations from 2 bar to 0-10 bar. During the variations, the void fraction depends on the Reynolds number of the flow and on the gas content of the water. The bulk modulus has been computed with different conditions. Most results concern pressures slightly over the vapor pressure. Air content has a strong influence on cavitation and on water compressibility after a vapor cavity collapse

  1. Cooling of pressurized water nuclear reactor vessels

    International Nuclear Information System (INIS)

    Curet, H.D.

    1978-01-01

    The improvement of pressurized water nuclear reactor vessels comprising flow dividers providing separate and distinct passages for the flow of core coolant water from each coolant water inlet, the flow dividers being vertically disposed in the annular flow areas provided by the walls of the vessel, the thermal shield (if present), and the core barrel is described. In the event of rupture of one of the coolant water inlet lines, water, especially emergency core coolant water, in the intact lines is thus prevented from by-passing the core by circumferential flow around the outermost surface of the core barrel and is instead directed so as to flow vertically downward through the annulus area between the vessel wall and the core barrel in a more normal manner to increase the probability of cooling of the core by the available cooling water in the lower plenum, thus preventing or delaying thermal damage to the core, and providing time for other appropriate remedial or damage preventing action by the operator

  2. Acceptance Test Report for the high pressure water jet system canister cleaning fixture

    Energy Technology Data Exchange (ETDEWEB)

    Burdin, J.R.

    1995-10-25

    This Acceptance Test confirmed the test results and recommendations, documented in WHC-SD-SNF-DTR-001, Rev. 0 Development Test Report for the High Pressure Water Jet System Nozzles, for decontaminating empty fuel canisters in KE-Basin. Optimum water pressure, water flow rate, nozzle size and overall configuration were tested

  3. Acceptance Test Report for the high pressure water jet system canister cleaning fixture

    International Nuclear Information System (INIS)

    Burdin, J.R.

    1995-01-01

    This Acceptance Test confirmed the test results and recommendations, documented in WHC-SD-SNF-DTR-001, Rev. 0 Development Test Report for the High Pressure Water Jet System Nozzles, for decontaminating empty fuel canisters in KE-Basin. Optimum water pressure, water flow rate, nozzle size and overall configuration were tested

  4. 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

  5. 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

  6. Pressure Dependence of the Liquid-Liquid Phase Transition of Nanopore Water Doped Slightly with Hydroxylamine, and a Phase Behavior Predicted for Pure Water

    Science.gov (United States)

    Nagoe, Atsushi; Iwaki, Shinji; Oguni, Masaharu; Tôzaki, Ken-ichi

    2014-09-01

    Phase transition behaviors of confined pure water and confined water doped with a small amount of hydroxylamine (HA) with a mole fraction of xHA = 0.03 were examined by high-pressure differential thermal analyses at 0.1, 50, 100, and 150 MPa; the average diameters of silica pores used were 2.0 and 2.5 nm. A liquid-liquid phase transition (LLPT) of the confined HA-doped water was clearly observed and its pressurization effect could be evaluated, unlike in the experiments on undoped water. It was found that pressurization causes the transition temperature (Ttrs) to linearly decrease, indicating that the low-temperature phase has a lower density than the high-temperature one. Transition enthalpy (ΔtrsH) decreased steeply with increasing pressure. Considering the linear decrease in Ttrs with increasing pressure, the steep decrease in ΔtrsH indicates that the LLPT effect of the HA-doped water attenuates with pressure. We present a new scenario of the phase behavior concerning the LLPT of pure water based on the analogy from the behavior of slightly HA-doped water, where a liquid-liquid critical point (LLCP) and a coexistence line are located in a negative-pressure regime but not in a positive-pressure one. It is reasonably understood that doping a small amount of HA into water results in negative chemical pressurization and causes the LLPT to occur even at ambient pressure.

  7. Water pressure and ground vibrations induced by water guns at a backwater pond on the Illinois River near Morris, Illinois

    Science.gov (United States)

    Koebel, Carolyn M.; Egly, Rachel M.

    2016-09-27

    Three different geophysical sensor types were used to characterize the underwater pressure waves and ground velocities generated by the underwater firing of seismic water guns. These studies evaluated the use of water guns as a tool to alter the movement of Asian carp. Asian carp are aquatic invasive species that threaten to move into the Great Lakes Basin from the Mississippi River Basin. Previous studies have identified a threshold of approximately 5 pounds per square inch (lb/in2) for behavioral modification and for structural limitation of a water gun barrier.Two studies were completed during August 2014 and May 2015 in a backwater pond connected to the Illinois River at a sand and gravel quarry near Morris, Illinois. The August 2014 study evaluated the performance of two 80-cubic-inch (in3) water guns. Data from the 80-in3 water guns showed that the pressure field had the highest pressures and greatest extent of the 5-lb/in2 target value at a depth of 5 feet (ft). The maximum recorded pressure was 13.7 lb/in2, approximately 25 ft from the guns. The produced pressure field took the shape of a north-south-oriented elongated sphere with the 5-lb/in2 target value extending across the entire study area at a depth of 5 ft. Ground velocities were consistent over time, at 0.0067 inches per second (in/s) in the transverse direction, 0.031 in/s in the longitudinal direction, and 0.013 in/s in the vertical direction.The May 2015 study evaluated the performance of one and two 100-in3 water guns. Data from the 100-in3 water guns, fired both individually and simultaneously, showed that the pressure field had the highest pressures and greatest extent of the 5-lb/in2 target value at a depth of 5 ft. The maximum pressure was 57.4 lb/in2, recorded at the underwater blast sensor closest to the water guns (at a horizontal distance of approximately 3 ft), as two guns fired simultaneously. Pressures and extent of the 5-lb/in2 target value decrease above and below this 5-ft depth

  8. A computer program for accident calculations of a standard pressurized water reactor

    International Nuclear Information System (INIS)

    Keutner, H.

    1979-01-01

    In this computer program the dynamic of a standard pressurized water reactor should be realized by both circulation loops with all important components. All important phenomena are taken into consideration, which appear for calculation of disturbances in order to state a realistic process for some minutes after a disturbance or a desired change of condition. In order to optimize the computer time simplifications are introduced in the statement of a differential-algebraic equalization system such that all important effects are taken into consideration. The model analysis starts from the heat production of the fuel rod via cladding material to the cooling medium water and considers the delay time from the core to the steam generator. Alternations of the cooling medium pressure as well as the different temperatures in the primary loop influence the pressuring system - the pressurizer - which is realized by a water and a steam zone with saturated and superheated steam respectively saturated and undercooled water with injection, heating and blow-down devices. The bilance of the steam generator to the secondary loop realizes the process engineering devices. Thereby the control regulation of the steam pressure and the reactor performance is realized. (orig.) [de

  9. European passive plant program A design for the 21st century

    International Nuclear Information System (INIS)

    Adomaitis, D.; Oyarzabal, M.

    1998-01-01

    In 1994, a group of European utilities initiated, together with Westinghouse and its industrial partner GENESI (an Italian consortium including ANSALDO and FIAT), a program designated EPP (European Passive Plant) to evaluate Westinghouse passive nuclear plant technology for application in Europe. The following major tasks were accomplished: (1) the impacts of the European utility requirements (EUR) on the Westinghouse nuclear island design were evaluated; and (2) a 1000 MWe passive plant reference design (EP1000) was established which conforms to the EUR and is expected to be licensable in Europe. With respect to safety systems and containment, the reference plant design closely follows that of the Westinghouse simplified pressurized water reactor (SPWR) design, while the AP600 plant design has been taken as the basis for the EP1000 reference design in the auxiliary system design areas. However, the EP1000 design also includes features required to meet the EUR, as well as key European licensing requirements. (orig.)

  10. Methodology of fuel rod design for pressurized light water reactors

    International Nuclear Information System (INIS)

    Teixeira e Silva, A.; Esteves, A.M.

    1988-01-01

    The fuel performance program FRAPCON-1 and the structural finite element program SAP-IV are applied in a pressurized water reactor fuel rod design methodology. The applied calculation procedure allows to dimension the fuel rod components and characterize its internal pressure. (author) [pt

  11. 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

  12. Effect of water pressure on absorbency of hydroentangled greige cotton nonwoven fabrics

    Science.gov (United States)

    A studied has been conducted to determine the effect of water pressure in a commercial-grade Fleissner MiniJet hydroentanglement system on the absorbency of greige (non-bleached) cotton lint-based nonwoven fabric. The study has shown that a water pressure of 125 Bar or higher on only two high-pressu...

  13. EDF's nuclear safety approach for pressurized water reactors

    International Nuclear Information System (INIS)

    Tanguy, P.; Kus, J.P.

    1987-01-01

    The realization of the important French program fifty-four units equipped with pressurized water reactors in service, or under construction-had led to the progressive implementation of an original approach in the field of nuclear safety. From an initial core consisting of the deterministic approach to safety devised on the other side of the Atlantic, which has been entirely preserved and often specified, further extras have been added which overall increase the level of safety of the installations, without any particular complications. This paper aims at presenting succinctly the outcome of the deliberation, which constitutes now the approach adopted by Electricite de France for the safety of nuclear units equipped with pressurized water reactors. This approach is explained in more detail in EDF's 'with book' on nuclear safety. (author)

  14. EDF'S nuclear safety approach for pressurized water reactors

    International Nuclear Information System (INIS)

    Tanguy, P.; Kus, J.P.

    1988-01-01

    The realization of the important French program fifty-four units equipped with pressurized water reactors in service, or under construction - had led to the progressive implementation of an original approach in the field of nuclear safety. From an initial core consisting of the deterministic approach to safety devised on the other side of the Atlantic, which has been entirely preserved and often specified, further extras have been added which overall increase the level of safety of the installations, without any particular complications. This paper aims at presenting succinctly the outcome of the deliberation, which constitutes now the approach adopted by Electricite de France for the safety of nuclear units equipped with pressurized water reactors. This approach is explained in more detail in EDF's white book on nuclear safety

  15. The self-similar turbulent flow of low-pressure water vapor

    Science.gov (United States)

    Konyukhov, V. K.; Stepanov, E. V.; Borisov, S. K.

    2018-05-01

    We studied turbulent flows of water vapor in a pipe connecting two closed vessels of equal volume. The vessel that served as a source of water vapor was filled with adsorbent in the form of corundum ceramic balls. These ceramic balls were used to obtain specific conditions to lower the vapor pressure in the source vessel that had been observed earlier. A second vessel, which served as a receiver, was empty of either air or vapor before each vapor sampling. The rate of the pressure increase in the receiver vessel was measured in a series of six samplings performed with high precision. The pressure reduction rate in the source vessel was found to be three times lower than the pressure growth rate in the receiver vessel. We found that the pressure growth rates in all of the adjacent pairs of samples could be arranged in a combination that appeared to be identical for all pairs, and this revealed the existence of a rather interesting and peculiar self-similarity law for the sampling processes under consideration.

  16. Development of plant dynamic analysis code for integrated self-pressurized water reactor (ISPDYN), and comparative study of pressure control methods

    International Nuclear Information System (INIS)

    Kusunoki, Tsuyoshi; Yokomura, Takeyoshi; Nabeshima, Kunihiko; Shimazaki, Junya; Shinohara, Yoshikuni.

    1988-01-01

    This report describes the development of plant dynamic analysis code (ISPDYN) for integrated self-pressurized water reactor, and comparative study of pressure control methods with this code. ISPDYN is developed for integrated self-pressurized water reactor, one of the trial design by JAERI. In the transient responses, the calculated results by ISPDYN are in good agreement with the DRUCK calculations. In addition, this report presents some sensitivity studies for selected cases. Computing time of this code is very short so as about one fifth of real time. The comparative study of self-pressurized system with forced-pressurized system by this code, for rapid load decrease and increase cases, has provided useful informations. (author)

  17. Application experience with ADBAC/DGH cationic surfactants for zebra mussel control in a nuclear service water system

    International Nuclear Information System (INIS)

    Jung, W.R.; Lacy, J.R.; Post, R.

    1992-01-01

    In response to the introduction and rapid growth of the zebra mussel population in the Great Lakes and the issuance of NRC Generic Letter 89-13 (Service Water Problems Affecting Safety-Related Equipment). A midwest nuclear station instituted a zebra mussel monitoring and control program. The nuclear station uses Lake Michigan as a cooling water source for two 1,100 MW Westinghouse 4-loop design, pressurized water reactors (PWR). Two years of monitoring indicated a growth in zebra mussel population from 0.5 organisms/m 2 in July 1990 to 100 organisms/m 2 by November 1990. This rapid increase indicated an urgent need for viable methods of zebra mussel control to protect the plant's essential service water (ESW) and non-essential service water (NESW) systems. In April 1991, the station formulated a plan that combined increased system inlet temperature with targeted application of a proprietary product containing two cationic surfactants, ADBAC/DGH. Sidestream biomonitoring boxes were seeded with zebra mussels and observed as a measure of the efficacy of the treatment. Where recommended dosages and duration were maintained, 100% control was achieved

  18. Evaporation of tungsten in vacuum at low hydrogen and water vapor pressures

    International Nuclear Information System (INIS)

    Andrievskij, R.A.; Galkin, E.A.; Khromonozhkin, V.V.

    1981-01-01

    The results of experimental investigations of tungsten evaporation rates in the temperature range 1650-2500 K, partial hydrogen and water vapours pressures 1x10 -5 -10 Pa are presented. Experi-- mental plant, equipment employed and radiometric technique of tungsten evaporation study are described. The dependences of evaporation rate and probabilities of tungsten oxidation by residual vacuum water vapours and dependences of tungsten evaporation rate on partial hydrogen and water vapours pressures are determined [ru

  19. Water-bearing, high-pressure Ca-silicates

    Science.gov (United States)

    Németh, Péter; Leinenweber, Kurt; Ohfuji, Hiroaki; Groy, Thomas; Domanik, Kenneth J.; Kovács, István J.; Kovács, Judit S.; Buseck, Peter R.

    2017-07-01

    Water-bearing minerals provide fundamental knowledge regarding the water budget of the mantle and are geophysically significant through their influence on the rheological and seismic properties of Earth's interior. Here we investigate the CaO-SiO2-H2O system at 17 GPa and 1773 K, corresponding to mantle transition-zone condition, report new high-pressure (HP) water-bearing Ca-silicates and reveal the structural complexity of these phases. We document the HP polymorph of hartrurite (Ca3SiO5), post-hartrurite, which is tetragonal with space group P4/ncc, a = 6.820 (5), c = 10.243 (8) Å, V = 476.4 (8) Å3, and Z = 4, and is isostructural with Sr3SiO5. Post-hartrurite occurs in hydrous and anhydrous forms and coexists with larnite (Ca2SiO4), which we find also has a hydrous counterpart. Si is 4-coordinated in both post-hartrurite and larnite. In their hydrous forms, H substitutes for Si (4H for each Si; hydrogrossular substitution). Fourier transform infrared (FTIR) spectroscopy shows broad hydroxyl absorption bands at ∼3550 cm-1 and at 3500-3550 cm-1 for hydrous post-hartrurite and hydrous larnite, respectively. Hydrous post-hartrurite has a defect composition of Ca2.663Si0.826O5H1.370 (5.84 weight % H2O) according to electron-probe microanalysis (EPMA), and the Si deficiency relative to Ca is also observed in the single-crystal data. Hydrous larnite has average composition of Ca1.924Si0.851O4H0.748 (4.06 weight % H2O) according to EPMA, and it is in agreement with the Si occupancy obtained using X-ray data collected on a single crystal. Superlattice reflections occur in electron-diffraction patterns of the hydrous larnite and could indicate crystallographic ordering of the hydroxyl groups and their associated cation defects. Although textural and EPMA-based compositional evidence suggests that hydrous perovskite may occur in high-Ca-containing (or low silica-activity) systems, the FTIR measurement does not show a well-defined hydroxyl absorption band for this

  20. Spreading pressures of water and n-propanol on polymer surfaces

    NARCIS (Netherlands)

    Busscher, H.J.; Kip, Gerhardus A.M.; van Silfhout, Arend; Arends, J.

    1986-01-01

    Spreading pressures of water and n-propanol on polytetrafluoroethylene (PTFE), polystyrene (PS), polymethylmethacrylate (PMMA), polycarbonate (PC), and glass are determined from ellipsometrically measured adsorption isotherms by graphical integration, yielding for water 9, 37, 26, 33, and 141

  1. Reducing energy consumption and leakage by active pressure control in a water supply system

    NARCIS (Netherlands)

    Bakker, M.; Rajewicz, T.; Kien, H.; Vreeburg, J.H.G.; Rietveld, L.C.

    2013-01-01

    WTP Gruszczyn supplies drinking water to a part of the city of Pozna?, in the Midwest of Poland. For the optimal automatic pressure control of the clear water pumping station, nine pressure measuring points were installed in the distribution network, and an active pressure control model was

  2. Bridge pressure flow scour for clear water conditions

    Science.gov (United States)

    2009-10-01

    The equilibrium scour at a bridge caused by pressure flow with critical approach velocity in clear-water simulation conditions was studied both analytically and experimentally. The flume experiments revealed that (1) the measured equilibrium scour pr...

  3. The influence of chemicals on water quality in a high pressure separation rig

    Energy Technology Data Exchange (ETDEWEB)

    Johnsen, Einar E.; Hemmingsen, Paal V.; Mediaas, Heidi; Svarstad, May Britt E.; Westvik, Arild

    2006-03-15

    In the research laboratory of Statoil at Rotvoll, Trondheim, a high pressure experimental rig used for separation and foaming studies has been developed. There have been several studies to ensure that the high pressure separation rig produces reliable and consistent results with regard to the water-in-oil and oil-in-water contents. The results are consistent with available field data and, just as important, consistent when changing variables like temperature, pressure drop and water cut. The results are also consistent when changing hydrodynamic variables like flow velocity and mixing point (using different choke valves) and when using oil with and without gas saturation. At equal experimental conditions, the high pressure separation rig is able to differentiate between separation characteristics of oil and water from different fields and from different wells at the same field. The high pressure separation and foam rig can be used from -10 deg C to 175 deg C and at pressures up to 200 bar. Crude oil and water are studied under relevant process conditions with respect to temperature, pressure, shear, water cut and separation time. In the present work the influence of chemicals on the oil and water quality has been studied. Chemicals have been mixed into the oil and/or water beforehand or added in situ (on-stream; simulated well stream). The amount of oil in the water after a given residence time in the separation cell has been measured. The results from the high pressure rig show that some demulsifiers, with their primary purpose of giving less water in oil, also have influence on the water quality. Improvement of water quality has been observed as well as no effect or aggravation. The experimental results have been compared to results from bottle tests at the field. The results from the bottle tests and from the laboratory are not corresponding, and only a full-scale field test can tell which of them are the correct results, if any. (Experience from corresponding

  4. 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

  5. Pressure wave propagation in the discharge piping with water pool

    International Nuclear Information System (INIS)

    Bang, Young S.; Seul, Kwang W.; Kim, In Goo

    2004-01-01

    Pressure wave propagation in the discharge piping with a sparger submerged in a water pool, following the opening of a safety relief valve, is analyzed. To predict the pressure transient behavior, a RELAP5/MOD3 code is used. The applicability of the RELAP5 code and the adequacy of the present modeling scheme are confirmed by simulating the applicable experiment on a water hammer with voiding. As a base case, the modeling scheme was used to calculate the wave propagation inside a vertical pipe with sparger holes and submerged within a water pool. In addition, the effects on wave propagation of geometric factors, such as the loss coefficient, the pipe configuration, and the subdivision of sparger pipe, are investigated. The effects of inflow conditions, such as water slug inflow and the slow opening of a safety relief valve are also examined

  6. The effects of pulse pressure from seismic water gun technology on Northern Pike

    Science.gov (United States)

    Gross, Jackson A.; Irvine, Kathryn M.; Wilmoth, Siri K.; Wagner, Tristany L.; Shields, Patrick A; Fox, Jeffrey R.

    2013-01-01

    We examined the efficacy of sound pressure pulses generated from a water gun for controlling invasive Northern Pike Esox lucius. Pulse pressures from two sizes of water guns were evaluated for their effects on individual fish placed at a predetermined random distance. Fish mortality from a 5,620.8-cm3 water gun (peak pressure source level = 252 dB referenced to 1 μP at 1 m) was assessed every 24 h for 168 h, and damage (intact, hematoma, or rupture) to the gas bladder, kidney, and liver was recorded. The experiment was replicated with a 1,966.4-cm3 water gun (peak pressure source level = 244 dB referenced to 1 μP at 1 m), but fish were euthanized immediately. The peak sound pressure level (SPLpeak), peak-to-peak sound pressure level (SPLp-p), and frequency spectrums were recorded, and the cumulative sound exposure level (SELcum) was subsequently calculated. The SPLpeak, SPLp-p, and SELcum were correlated, and values varied significantly by treatment group for both guns. Mortality increased and organ damage was greater with decreasing distance to the water gun. Mortality (31%) by 168 h was only observed for Northern Pike exhibiting the highest degree of organ damage. Mortality at 72 h and 168 h postexposure was associated with increasing SELcum above 195 dB. The minimum SELcum calculated for gas bladder rupture was 199 dB recorded at 9 m from the 5,620.8-cm3 water gun and 194 dB recorded at 6 m from the 1,966.4-cm3water gun. Among Northern Pike that were exposed to the large water gun, 100% of fish exposed at 3 and 6 m had ruptured gas bladders, and 86% exposed at 9 m had ruptured gas bladders. Among fish that were exposed to pulse pressures from the smaller water gun, 78% exhibited gas bladder rupture. Results from these initial controlled experiments underscore the potential of water guns as a tool for controlling Northern Pike.

  7. Effects of ambient temperature and water vapor on chamber pressure and oxygen level during low atmospheric pressure stunning of poultry.

    Science.gov (United States)

    Holloway, Paul H; Pritchard, David G

    2017-08-01

    The characteristics of the vacuum used in a low atmospheric pressure stunning system to stun (render unconscious) poultry prior to slaughter are described. A vacuum chamber is pumped by a wet screw compressor. The vacuum pressure is reduced from ambient atmospheric pressure to an absolute vacuum pressure of ∼250 Torr (∼33 kPa) in ∼67 sec with the vacuum gate valve fully open. At ∼250 Torr, the sliding gate valve is partially closed to reduce effective pumping speed, resulting in a slower rate of decreasing pressure. Ambient temperature affects air density and water vapor pressure and thereby oxygen levels and the time at the minimum total pressure of ∼160 Torr (∼21 kPa) is varied from ∼120 to ∼220 sec to ensure an effective stun within the 280 seconds of each cycle. The reduction in total pressure results in a gradual reduction of oxygen partial pressure that was measured by a solid-state electrochemical oxygen sensor. The reduced oxygen pressure leads to hypoxia, which is recognized as a humane method of stunning poultry. The system maintains an oxygen concentration of air always reduces the oxygen concentrations to a value lower than in dry air. The partial pressure of water and oxygen were found to depend on the pump down parameters due to the formation of fog in the chamber and desorption of water from the birds and the walls of the vacuum chamber. © The Author 2017. Published by Oxford University Press on behalf of Poultry Science Association.

  8. Steam explosions-induced containment failure studies for Swiss nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Zuchuat, O.; Schmocker, U. [Swiss Federal Nuclear Safety Inspectorate, Villigen (Switzerland); Esmaili, H.; Khatib-Rahbar, M.

    1998-01-01

    The assessment of the consequences of both in-vessel and ex-vessel energetic fuel-coolant interaction for Beznau (a Westinghouse pressurized water reactor with a large, dry containment), Goesgen (a Siemens/KWU pressurized water reactor with a large, dry containment) and Leibstadt (a General Electric boiling water reactor-6 with a free standing steel, MARK-III containment) nuclear power plants is presented in this paper. The Conditional Containment Failure Probability of the steel containment of these Swiss nuclear power plants is determined based on different probabilistic approaches. (author)

  9. Effect of Leaf Water Potential on Internal Humidity and CO2 Dissolution: Reverse Transpiration and Improved Water Use Efficiency under Negative Pressure.

    Science.gov (United States)

    Vesala, Timo; Sevanto, Sanna; Grönholm, Tiia; Salmon, Yann; Nikinmaa, Eero; Hari, Pertti; Hölttä, Teemu

    2017-01-01

    The pull of water from the soil to the leaves causes water in the transpiration stream to be under negative pressure decreasing the water potential below zero. The osmotic concentration also contributes to the decrease in leaf water potential but with much lesser extent. Thus, the surface tension force is approximately balanced by a force induced by negative water potential resulting in concavely curved water-air interfaces in leaves. The lowered water potential causes a reduction in the equilibrium water vapor pressure in internal (sub-stomatal/intercellular) cavities in relation to that over water with the potential of zero, i.e., over the flat surface. The curved surface causes a reduction also in the equilibrium vapor pressure of dissolved CO 2 , thus enhancing its physical solubility to water. Although the water vapor reduction is acknowledged by plant physiologists its consequences for water vapor exchange at low water potential values have received very little attention. Consequences of the enhanced CO 2 solubility to a leaf water-carbon budget have not been considered at all before this study. We use theoretical calculations and modeling to show how the reduction in the vapor pressures affects transpiration and carbon assimilation rates. Our results indicate that the reduction in vapor pressures of water and CO 2 could enhance plant water use efficiency up to about 10% at a leaf water potential of -2 MPa, and much more when water potential decreases further. The low water potential allows for a direct stomatal water vapor uptake from the ambient air even at sub-100% relative humidity values. This alone could explain the observed rates of foliar water uptake by e.g., the coastal redwood in the fog belt region of coastal California provided the stomata are sufficiently open. The omission of the reduction in the water vapor pressure causes a bias in the estimates of the stomatal conductance and leaf internal CO 2 concentration based on leaf gas exchange

  10. 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)

  11. Molecular Dynamics of Equilibrium and Pressure-Driven Transport Properties of Water through LTA-Type Zeolites

    KAUST Repository

    Turgman-Cohen, Salomon; Araque, Juan C.; Hoek, Eric M. V.; Escobedo, Fernando A.

    2013-01-01

    We consider an atomistic model to investigate the flux of water through thin Linde type A (LTA) zeolite membranes with differing surface chemistries. Using molecular dynamics, we have studied the flow of water under hydrostatic pressure through a fully hydrated LTA zeolite film (∼2.5 nm thick) capped with hydrophilic and hydrophobic moieties. Pressure drops in the 50-400 MPa range were applied across the membrane, and the flux of water was monitored for at least 15 ns of simulation time. For hydrophilic membranes, water molecules adsorb at the zeolite surface, creating a highly structured fluid layer. For hydrophobic membranes, a depletion of water molecules occurs near the water/zeolite interface. For both types of membranes, the water structure is independent of the pressure drop established in the system and the flux through the membranes is lower than that observed for the bulk zeolitic material; the latter allows an estimation of surface barrier effects to pressure-driven water transport. Mechanistically, it is observed that (i) bottlenecks form at the windows of the zeolite structure, preventing the free flow of water through the porous membrane, (ii) water molecules do not move through a cage in a single-file fashion but rather exhibit a broad range of residence times and pronounced mixing, and (iii) a periodic buildup of a pressure difference between inlet and outlet cages takes place which leads to the preferential flow of water molecules toward the low-pressure cages. © 2013 American Chemical Society.

  12. Molecular Dynamics of Equilibrium and Pressure-Driven Transport Properties of Water through LTA-Type Zeolites

    KAUST Repository

    Turgman-Cohen, Salomon

    2013-10-08

    We consider an atomistic model to investigate the flux of water through thin Linde type A (LTA) zeolite membranes with differing surface chemistries. Using molecular dynamics, we have studied the flow of water under hydrostatic pressure through a fully hydrated LTA zeolite film (∼2.5 nm thick) capped with hydrophilic and hydrophobic moieties. Pressure drops in the 50-400 MPa range were applied across the membrane, and the flux of water was monitored for at least 15 ns of simulation time. For hydrophilic membranes, water molecules adsorb at the zeolite surface, creating a highly structured fluid layer. For hydrophobic membranes, a depletion of water molecules occurs near the water/zeolite interface. For both types of membranes, the water structure is independent of the pressure drop established in the system and the flux through the membranes is lower than that observed for the bulk zeolitic material; the latter allows an estimation of surface barrier effects to pressure-driven water transport. Mechanistically, it is observed that (i) bottlenecks form at the windows of the zeolite structure, preventing the free flow of water through the porous membrane, (ii) water molecules do not move through a cage in a single-file fashion but rather exhibit a broad range of residence times and pronounced mixing, and (iii) a periodic buildup of a pressure difference between inlet and outlet cages takes place which leads to the preferential flow of water molecules toward the low-pressure cages. © 2013 American Chemical Society.

  13. Effects of High Hydrostatic Pressure on Water Absorption of Adzuki Beans

    Science.gov (United States)

    Ueno, Shigeaki; Shigematsu, Toru; Karo, Mineko; Hayashi, Mayumi; Fujii, Tomoyuki

    2015-01-01

    The effect of high hydrostatic pressure (HHP) treatment on dried soybean, adzuki bean, and kintoki kidney bean, which are low-moisture-content cellular biological materials, was investigated from the viewpoint of water absorption. The samples were vacuum-packed with distilled water and pressurized at 200 MPa and 25 °C for 10 min. After the HHP treatment, time courses of the moisture contents of the samples were measured, and the dimensionless moisture contents were estimated. Water absorption in the case of soybean could be fitted well by a simple water diffusion model. High pressures were found to have negligible effects on water absorption into the cotyledon of soybean and kintoki kidney bean. A non-linear least square method based on the Weibull equation was applied for the adzuki beans, and the effective water diffusion coefficient was found to increase significantly from 8.6 × 10−13 to 6.7 × 10−10 m2/s after HHP treatment. Approximately 30% of the testa of the adzuki bean was damaged upon HHP treatment, which was comparable to the surface area of the testa in the partially peeled adzuki bean sample. Thus, HHP was confirmed to promote mass transfer to the cotyledon of legumes with a tight testa. PMID:28231195

  14. Effects of High Hydrostatic Pressure on Water Absorption of Adzuki Beans

    Directory of Open Access Journals (Sweden)

    Shigeaki Ueno

    2015-05-01

    Full Text Available The effect of high hydrostatic pressure (HHP treatment on dried soybean, adzuki bean, and kintoki kidney bean, which are low-moisture-content cellular biological materials, was investigated from the viewpoint of water absorption. The samples were vacuum-packed with distilled water and pressurized at 200 MPa and 25 °C for 10 min. After the HHP treatment, time courses of the moisture contents of the samples were measured, and the dimensionless moisture contents were estimated. Water absorption in the case of soybean could be fitted well by a simple water diffusion model. High pressures were found to have negligible effects on water absorption into the cotyledon of soybean and kintoki kidney bean. A non-linear least square method based on the Weibull equation was applied for the adzuki beans, and the effective water diffusion coefficient was found to increase significantly from 8.6 × 10−13 to 6.7 × 10−10 m2/s after HHP treatment. Approximately 30% of the testa of the adzuki bean was damaged upon HHP treatment, which was comparable to the surface area of the testa in the partially peeled adzuki bean sample. Thus, HHP was confirmed to promote mass transfer to the cotyledon of legumes with a tight testa.

  15. High protein flexibility and reduced hydration water dynamics are key pressure adaptive strategies in prokaryotes

    KAUST Repository

    Martinez, N.

    2016-09-06

    Water and protein dynamics on a nanometer scale were measured by quasi-elastic neutron scattering in the piezophile archaeon Thermococcus barophilus and the closely related pressure-sensitive Thermococcus kodakarensis, at 0.1 and 40 MPa. We show that cells of the pressure sensitive organism exhibit higher intrinsic stability. Both the hydration water dynamics and the fast protein and lipid dynamics are reduced under pressure. In contrast, the proteome of T. barophilus is more pressure sensitive than that of T. kodakarensis. The diffusion coefficient of hydration water is reduced, while the fast protein and lipid dynamics are slightly enhanced with increasing pressure. These findings show that the coupling between hydration water and cellular constituents might not be simply a master-slave relationship. We propose that the high flexibility of the T. barophilus proteome associated with reduced hydration water may be the keys to the molecular adaptation of the cells to high hydrostatic pressure.

  16. High protein flexibility and reduced hydration water dynamics are key pressure adaptive strategies in prokaryotes

    KAUST Repository

    Martinez, N.; Michoud, Gregoire; Cario, A.; Ollivier, J.; Franzetti, B.; Jebbar, M.; Oger, P.; Peters, J.

    2016-01-01

    Water and protein dynamics on a nanometer scale were measured by quasi-elastic neutron scattering in the piezophile archaeon Thermococcus barophilus and the closely related pressure-sensitive Thermococcus kodakarensis, at 0.1 and 40 MPa. We show that cells of the pressure sensitive organism exhibit higher intrinsic stability. Both the hydration water dynamics and the fast protein and lipid dynamics are reduced under pressure. In contrast, the proteome of T. barophilus is more pressure sensitive than that of T. kodakarensis. The diffusion coefficient of hydration water is reduced, while the fast protein and lipid dynamics are slightly enhanced with increasing pressure. These findings show that the coupling between hydration water and cellular constituents might not be simply a master-slave relationship. We propose that the high flexibility of the T. barophilus proteome associated with reduced hydration water may be the keys to the molecular adaptation of the cells to high hydrostatic pressure.

  17. Pressurized Fluidized Bed Combustion Second-Generation System Research and Development

    Energy Technology Data Exchange (ETDEWEB)

    A. Robertson; D. Horazak; R. Newby; H. Goldstein

    2002-11-01

    Research is being conducted under United States Department of Energy (DOE) Contract DE-AC21-86MC21023 to develop a new type of coal-fired plant for electric power generation. This new type of plant--called a Second-Generation or Advanced Pressurized Circulating Fluidized Bed Combustion (APCFB) plant--offers the promise of efficiencies greater than 45% (HHV), with both emissions and a cost of electricity that are significantly lower than conventional pulverized-coal-fired plants with scrubbers. The APCFB plant incorporates the partial gasification of coal in a carbonizer, the combustion of carbonizer char in a pressurized circulating fluidized bed boiler (PCFB), and the combustion of carbonizer syngas in a topping combustor to achieve gas turbine inlet temperatures of 2300 F and higher. A conceptual design was previously prepared for this new type of plant and an economic analysis presented, all based on the use of a Siemens Westinghouse W501F gas turbine with projected carbonizer, PCFB, and topping combustor performance data. Having tested these components at the pilot plant stage, the referenced conceptual design is being updated to reflect more accurate performance predictions together with the use of the more advanced Siemens Westinghouse W501G gas turbine and a conventional 2400 psig/1050 F/1050 F/2-1/2 in. steam turbine. This report describes the updated plant which is projected to have an HHV efficiency of 48% and identifies work completed for the October 2001 through September 2002 time period.

  18. Water solubility of synthetic pyrope at high temperature and pressure up to 12GPa

    Science.gov (United States)

    Huang, S.; Chen, J.

    2012-12-01

    Water can be incorporated into normally anhydrous minerals as OH- defects and transported into the mantle. Its existence in the mantle may affect property of minerals, such as elasticity, electrical conductivity and rheological properties. As the secondary mineral in the mantle, garnet has not been extensively studied for its water solubility and there is discrepancies among the existing experiments on the water solubility in the garnet change at pressures and temperatures. Geiger et al., 1991 investigated water content in synthetic pyrope and concluded 0.02wt% to 0.07wt% OH- substitution. Lu et al., 1997 found 198ppm water in the Dora Miara pyrope at 100Kbar and 1000°C. Withers et al., 1998 claimed that water solubility in pyrope reached 1000ppm at 5GPa and then decreased as pressure increasing; above 7GPa, no water was detected. Mookherjee et al., 2009 also explored pyrope-rich garnet, which contains water up to 0.1%wt at 5-9GPa and temperatures 1373K-1473K. Here we report a study of water solubility of synthetic single crystal pyrope at pressures 4-12GPa and temperature 1000°C. Single crystals of pyrope were synthesized using multi-anvil press and water contents in these samples were measured using FTIR. We have observed OH- peak at 3650 cm-1 along this pressure range, although Withers, 1998 reported water contents decrease to undetectable level above 7GPa. Water solubility in pyrope will be reported as a function of pressure up to 12 GPa at 1000°C.

  19. Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

    International Nuclear Information System (INIS)

    Greneche, D.

    2014-01-01

    This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

  20. 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)

  1. Fissures in rock under water pressure, implications on stability : 3 unusual cases

    Energy Technology Data Exchange (ETDEWEB)

    Helwig, P.C. [Helwig Hydrotechnique Ltd., St. John' s, NL (Canada)

    2006-07-01

    The presence of water in rock joints has important implications on the stability of rock foundations. Appropriate analyses are needed to assess the stability of dam foundations, abutments and rock walls. This paper presented 3 case studies in which the freezing of seepage flows in rock joints and transient pressure in rock walls were investigated: (1) an assessment of the effects of freezing water in rock joints at the Paradise River arch dam in Newfoundland; (2) stability of rock walls in the unlined power tunnel of the Cat Arm hydroelectric development in Newfoundland due to transient pressures; and (3) assessing the influence of fluctuating water pressures in a stilling basin excavated in rock. After an investigation of the Paradise River canyon walls, a drainage system comprised of peripheral drain holes was drilled into the foundation and walls at regular intervals to intercept seepage flows and to relieve uplift water pressures. However, no special treatment was found for the potential freezing of water in the joints of the dam walls and foundation. The Cat Arm tunnel was used to study the depth at which significant transient pressures can be used to assess rock stability. Rock properties, typical fracture apertures and spacing were assumed and joint deformability was taken into account. An axisymmetric solution was obtained by considering the continuity and flow through an annular element of the rock wall. A finite difference method was used to solve the resulting nonlinear differential equation. In the final case study, blast-damaged rock was undermining the toe of a spillway. A cut-off wall was constructed as a series of drilled, cast-in-place concrete caisson piles. Criteria for the design included extending the cut-off wall to a depth beyond the effects of fluctuating surface pressures. Depth was assessed by considering the transient behaviour of water penetrating a sub-vertical joint subject exposed to fluctuating pressures. Results of the calculations

  2. Fresh Water Generation from Aquifer-Pressured Carbon Storage: Annual Report FY09

    Energy Technology Data Exchange (ETDEWEB)

    Wolery, T; Aines, R; Hao, Y; Bourcier, W; Wolfe, T; Haussman, C

    2009-11-25

    This project is establishing the potential for using brine pressurized by Carbon Capture and Storage (CCS) operations in saline formations as the feedstock for desalination and water treatment technologies including reverse osmosis (RO) and nanofiltration (NF). The aquifer pressure resulting from the energy required to inject the carbon dioxide provides all or part of the inlet pressure for the desalination system. Residual brine is reinjected into the formation at net volume reduction, such that the volume of fresh water extracted balances the volume of CO{sub 2} injected into the formation. This process provides additional CO{sub 2} storage capacity in the aquifer, reduces operational risks (cap-rock fracturing, contamination of neighboring fresh water aquifers, and seismicity) by relieving overpressure in the formation, and provides a source of low-cost fresh water to offset costs or operational water needs. This multi-faceted project combines elements of geochemistry, reservoir engineering, and water treatment engineering. The range of saline formation waters is being identified and analyzed. Computer modeling and laboratory-scale experimentation are being used to examine mineral scaling and osmotic pressure limitations. Computer modeling is being used to evaluate processes in the storage aquifer, including the evolution of the pressure field. Water treatment costs are being evaluated by comparing the necessary process facilities to those in common use for seawater RO. There are presently limited brine composition data available for actual CCS sites by the site operators including in the U.S. the seven regional Carbon Sequestration Partnerships (CSPs). To work around this, we are building a 'catalog' of compositions representative of 'produced' waters (waters produced in the course of seeking or producing oil and gas), to which we are adding data from actual CCS sites as they become available. Produced waters comprise the most common

  3. Pressure Drop Test of Hybrid Mixing Vane Spacer Grid

    Energy Technology Data Exchange (ETDEWEB)

    Oh, D. S.; Chang, S. K.; Kim, B. D.; Chun, S. Y.; Chun, T. H

    2007-08-15

    The pressure loss test has been accomplished in the test section containing 5x5 rod bundle with a length of 2 m including 3 spacer grids. The test has been performed for the 5 kinds of spacer grids to compare the pressure loss characteristics: 1. Plain spacer grid which has the same body of the Hybrid but without vane (Plain), 2. Hybrid Vane spacer grid (Hybrid), 3. Hybrid-SC spacer grid which is constructed with coined, chamfered strip and is fabricated by spot welding, 4. Hybrid-LC spacer grid which is constructed with coined, chamfered strip and is fabricated by line welding along intersection line, 5. Westinghouse spacer grid with split vane (Plus-7). The pressure loss coefficient of the Plain, Hybrid, Hybrid-SC, Hybrid-LC, and Plus-7 spacer grid is 0.93, 1.15, 1.02, 1.04, and 1.08, respectively.

  4. Design of an integral missile shield in integrated head assembly for pressurized water reactor at commercial nuclear plants

    International Nuclear Information System (INIS)

    Baliga, Ravi; Watts, Tom Neal; Kamath, Harish

    2015-01-01

    containment and damaging other safety components in the containment that are required to be operated for safe shutdown of the reactor. The missile shield in the IHA is designed to absorb missile energy due to an impact from missiles associated with a postulated CRDM housing break. This paper provides details of the CRDM missile shield design in the IHA for Westinghouse Pressurized Water Reactors (PWR) and it can be extended to other PWRs such as VVERs. (author)

  5. Weak interactions between water and clathrate-forming gases at low pressures

    Energy Technology Data Exchange (ETDEWEB)

    Thurmer, Konrad; Yuan, Chunqing; Kimmel, Gregory A.; Kay, Bruce D.; Smith, R. Scott

    2015-11-01

    Using scanning probe microscopy and temperature programed desorption we examined the interaction between water and two common clathrate-forming gases, methane and isobutane, at low temperature and low pressure. Water co-deposited with up to 10-1 mbar methane or 10-5 mbar isobutane at 140 K onto a Pt(111) substrate yielded pure crystalline ice, i.e., the exposure to up to ~107 gas molecules for each deposited water molecule did not have any detectable effect on the growing films. Exposing metastable, less than 2 molecular layers thick, water films to 10-5 mbar methane does not alter their morphology, suggesting that the presence of the Pt(111) surface is not a strong driver for hydrate formation. This weak water-gas interaction at low pressures is supported by our thermal desorption measurements from amorphous solid water and crystalline ice where 1 ML of methane desorbs near ~43 K and isobutane desorbs near ~100 K. Similar desorption temperatures were observed for desorption from amorphous solid water.

  6. Assessment of Loss-of-Coolant Effect on Pressurized Heavy Water Reactors

    International Nuclear Information System (INIS)

    Kim, Won Young; Park, Joo Hwan; Kim, Bong Ghi

    2009-01-01

    A CANDU reactor is a heavy-water-moderated, natural uranium fuelled reactor with a pressure tube. The reactor contains a horizontal cylindrical vessel (calandria) and each pressure tube is isolated from the heavy-water moderator in a calandria. This allows the moderator system to be operated of a high-pressure and of a high-temperature coolant in pressure tube. This causes the pressurized liquid coolant in the channel to void and therefore give rise to a reactivity transient in the event of a break or fault in the coolant circuit. In particular, all CANDU reactors are well known to have a positive void reactivity coefficient and thus this phenomenon may lead to a positive feedback, which can cause a large power pulse. We assess the loss-of-coolant effect by coolant void reactivity versus fuel burnup, four factor parameters for fresh fuel and equilibrium fuel, reactivity change due to the change of coolant density and reactivity change in the case of half- and full-core coolant

  7. Additional Stress And Fracture Mechanics Analyses Of Pressurized Water Reactor Pressure Vessel Nozzles

    International Nuclear Information System (INIS)

    Walter, Matthew; Yin, Shengjun; Stevens, Gary; Sommerville, Daniel; Palm, Nathan; Heinecke, Carol

    2012-01-01

    In past years, the authors have undertaken various studies of nozzles in both boiling water reactors (BWRs) and pressurized water reactors (PWRs) located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Those studies described stress and fracture mechanics analyses performed to assess various RPV nozzle geometries, which were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-life (EOL) to require evaluation of embrittlement as part of the RPV analyses associated with pressure-temperature (P-T) limits. In this paper, additional stress and fracture analyses are summarized that were performed for additional PWR nozzles with the following objectives: To expand the population of PWR nozzle configurations evaluated, which was limited in the previous work to just two nozzles (one inlet and one outlet nozzle). To model and understand differences in stress results obtained for an internal pressure load case using a two-dimensional (2-D) axi-symmetric finite element model (FEM) vs. a three-dimensional (3-D) FEM for these PWR nozzles. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated. To investigate the applicability of previously recommended linear elastic fracture mechanics (LEFM) hand solutions for calculating the Mode I stress intensity factor for a postulated nozzle corner crack for pressure loading for these PWR nozzles. These analyses were performed to further expand earlier work completed to support potential revision and refinement of Title 10 to the U.S. Code of Federal Regulations (CFR), Part 50, Appendix G, Fracture Toughness Requirements, and are intended to supplement similar evaluation of nozzles presented at the 2008, 2009, and 2011 Pressure Vessels and Piping (PVP

  8. Design of a pressurized water loop heated by electric resistances

    International Nuclear Information System (INIS)

    Ribeiro, S.V.G.

    1981-01-01

    A pressurized water loop design is presented. Its operating pressure is 420 psi and we seek to simulate qualitatively some thermo-hydraulic phenomena of PWR reactors. The primary circuit simulator consists basically of two elements: 1)the test section housing 16 electric resistences dissipating a total power of 100 Kw; 2)the loop built of SCH40S 304L steel piping, consisting of the pump, a heat exchanger and the pressurizer. (Author) [pt

  9. 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

  10. Integral Pressurized Water Reactor Simulator Manual

    International Nuclear Information System (INIS)

    2017-01-01

    This publication provides detailed explanations of the theoretical concepts that the simulator users have to know to gain a comprehensive understanding of the physics and technology of integral pressurized water reactors. It provides explanations of each of the simulator screens and various controls that a user can monitor and modify. A complete description of all the simulator features is also provided. A detailed set of exercises is provided in the Exercise Handbook accompanying this publication.

  11. RELAP5-3D Code for Supercritical-Pressure Light-Water-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Riemke, Richard Allan; Davis, Cliff Bybee; Schultz, Richard Raphael

    2003-04-01

    The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point.

  12. 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

  13. Probabilistic integrity assessment of pressure tubes in an operating pressurized heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Young-Jin; Park, Heung-Bae [KEPCO E and C, 188 Gumi-dong, Bundang-gu, Seongnam-si, Gyeonggi-do 463-870 (Korea, Republic of); Lee, Jung-Min; Kim, Young-Jin [School of Mechanical Engineering, Sungkyunkwan University, 300 Chunchun-dong, Jangan-gu, Suwon-si, Gyeonggi-do 440-746 (Korea, Republic of); Ko, Han-Ok [Korea Institute of Nuclear Safety, 34 Gwahak-ro, Yuseong-gu, Daejeon-si 305-338 (Korea, Republic of); Chang, Yoon-Suk, E-mail: yschang@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, 1 Seocheon-dong, Giheung-gu, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of)

    2012-02-15

    Even though pressure tubes are major components of a pressurized heavy water reactor (PHWR), only small proportions of pressure tubes are sampled for inspection due to limited inspection time and costs. Since the inspection scope and integrity evaluation have been treated by using a deterministic approach in general, a set of conservative data was used instead of all known information related to in-service degradation mechanisms because of inherent uncertainties in the examination. Recently, in order that pressure tube degradations identified in a sample of inspected pressure tubes are taken into account to address the balance of the uninspected ones in the reactor core, a probabilistic approach has been introduced. In the present paper, probabilistic integrity assessments of PHWR pressure tubes were carried out based on accumulated operating experiences and enhanced technology. Parametric analyses on key variables were conducted, which were periodically measured by in-service inspection program, such as deuterium uptake rate, dimensional change rate of pressure tube and flaw size distribution. Subsequently, a methodology to decide optimum statistical distribution by using a robust method adopting a genetic algorithm was proposed and applied to the most influential variable to verify the reliability of the proposed method. Finally, pros and cons of the alternative distributions comparing with corresponding ones derived from the traditional method as well as technical findings from the statistical assessment were discussed to show applicability to the probabilistic assessment of pressure tubes.

  14. Recent developments in high pressure water technology

    International Nuclear Information System (INIS)

    Johnson, N.A.; Johnson, T.

    1992-01-01

    High Pressure Water Jetting has advanced rapidly in the last decade to a point where the field is splitting into specialised areas. This has left the end user or client in the dark as to whether water jetting will work and if so what equipment is best suited to their particular application. The aim of this paper is to give an overview of:-1. The way water is delivered to the surface and the parameters which control the concentration of energy available on impact. 2. The factors governing application driven selection of equipment. 3. The effects to technical advances in pumps and delivery systems on equipment selection with reference to their to their application to concrete removal and nuclear decontamination. (Author)

  15. Importance of pressure reducing valves (PRVs) in water supply networks.

    Science.gov (United States)

    Signoreti, R. O. S.; Camargo, R. Z.; Canno, L. M.; Pires, M. S. G.; Ribeiro, L. C. L. J.

    2016-08-01

    Challenged with the high rate of leakage from water supply systems, these managers are committed to identify control mechanisms. In order to standardize and control the pressure Pressure Reducing Valves (VRP) are installed in the supply network, shown to be more effective and provide a faster return for the actual loss control measures. It is known that the control pressure is while controlling the occurrence of leakage. Usually the network is sectored in areas defined by pressure levels according to its topography, once inserted the VRP in the same system will limit the downstream pressure. This work aims to show the importance of VRP as loss reduction for tool.

  16. 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

  17. Water permeability of nanoporous graphene at realistic pressures for reverse osmosis desalination

    Energy Technology Data Exchange (ETDEWEB)

    Cohen-Tanugi, David; Grossman, Jeffrey C. [Department of Materials Science and Engineering, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States)

    2014-08-21

    Nanoporous graphene (NPG) shows tremendous promise as an ultra-permeable membrane for water desalination thanks to its atomic thickness and precise sieving properties. However, a significant gap exists in the literature between the ideal conditions assumed for NPG desalination and the physical environment inherent to reverse osmosis (RO) systems. In particular, the water permeability of NPG has been calculated previously based on very high pressures (1000–2000 bars). Does NPG maintain its ultrahigh water permeability under real-world RO pressures (<100 bars)? Here, we answer this question by drawing results from molecular dynamics simulations. Our results indicate that NPG maintains its ultrahigh permeability even at low pressures, allowing a permeate water flux of 6.0 l/h-bar per pore, or equivalently 1041 ± 20 l/m{sup 2}-h-bar assuming a nanopore density of 1.7 × 10{sup 13} cm{sup −2}.

  18. Fracture risk assessment for the pressurized water reactor pressure vessel under pressurized thermal shock events

    International Nuclear Information System (INIS)

    Chou, Hsoung-Wei; Huang, Chin-Cheng

    2016-01-01

    Highlight: • The PTS loading conditions consistent with the USNRC's new PTS rule are applied as the loading condition for a Taiwan domestic PWR. • The state-of-the-art PFM technique is employed to analyze a reactor pressure vessel. • Novel flaw model and embrittlement correlation are considered in the study. • The RT-based regression formula of NUREG-1874 was also utilized to evaluate the failure risks of RPV. • For slightly embrittled RPV, the SO-1 type PTSs play more important role than other types of PTS. - Abstract: The fracture risk of the pressurized water reactor pressure vessel of a Taiwan domestic nuclear power plant has been evaluated according to the technical basis of the U.S.NRC's new pressurized thermal shock (PTS) screening criteria. The ORNL's FAVOR code and the PNNL's flaw models were employed to perform the probabilistic fracture mechanics analysis associated with plant specific parameters of the domestic reactor pressure vessel. Meanwhile, the PTS thermal hydraulic and probabilistic risk assessment data analyzed from a similar nuclear power plant in the United States for establishing the new PTS rule were applied as the loading conditions. Besides, an RT-based regression formula derived by the U.S.NRC was also utilized to verify the through-wall cracking frequencies. It is found that the through-wall cracking of the analyzed reactor pressure vessel only occurs during the PTS events resulted from the stuck-open primary safety relief valves that later reclose, but with only an insignificant failure risk. The results indicate that the Taiwan domestic PWR pressure vessel has sufficient structural margin for the PTS attack until either the current license expiration dates or during the proposed extended operation periods.

  19. Reuse of waste water from high pressure water jet decontamination for reactor decommissioning scrap metal

    International Nuclear Information System (INIS)

    Deng Junxian; Li Xin; Hou Huijuan

    2011-01-01

    For recycle and reuse of reactor decommissioning scrap metal by high pressure water jet decontamination, large quantity of radioactive waste water will be generated. To save the cost of radioactive waste water treatment and to reduce the cost of the scrap decontamination, this part of radioactive waste water should be reused. Most of the radioactivities in the decontamination waste water come from the solid particle in the water. Thus to reuse the waste water, the solid particle in the waster should be removed. Different possible treatment technologies have been investigated. By cost benefit analysis the centrifugal separation technology is selected. (authors)

  20. Assessment of EPRI water chemistry guidelines for new nuclear power plants

    International Nuclear Information System (INIS)

    Reid Richard; Kim Karen; McCree, Anisa; Eaker, Richard; Sawochka, Steve; Giannelli, Joe

    2012-09-01

    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 currently 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 state-of-the-art, industry developed water chemistry controls. In parallel, the industry will need to consider and update water chemistry guidelines as well as plant startup and operational strategies based on the advanced plant designs. EPRI has performed assessments of water chemistry control guidance or assumptions provided in design and licensing documents for several advanced plant designs. These designs include: Westinghouse AP1000 Pressurized Water Reactor AREVA US-EPR Pressurized Water Reactor Mitsubishi Nuclear Energy Systems/Mitsubishi Heavy Industries Advanced Pressurized Water Reactor Korea Hydro and Nuclear Power APR1400 Pressurized Water Reactor Toshiba Advanced Boiling Water Reactor (ABWR) General Electric-Hitachi Economic Simplified Boiling Water Reactor (ESBWR) The intent of these assessments was to identify key design differences in each of the new plant designs relative to the current operating fleet and to identify differences in water chemistry specifications or design assumptions provided in design and licensing documents for the plants in comparison to current EPRI Water Chemistry Guidelines. This paper provides a summary of the key results of these assessments. The fundamental design and operation of the advanced plants is similar to the currently operating fleet. As such, the new plants are

  1. Development of alternative fuel for pressurized water reactors

    International Nuclear Information System (INIS)

    Cardoso, P.E.; Ferreira, R.A.N.; Ferraz, W.B.; Lameiras, F.S.; Santos, A.; Assis, G. de; Doerr, W.O.; Wehner, E.L.

    1984-01-01

    The utilization of alternative fuel cycles in Pressurized Water Reactors (PWR) such as Th/U and Th/Pu cycles can permit a better utilization of uranium reserves without the necessity of developing new power reactor concepts. The development of the technology of alternative fuels for PWR is one of the objectives of the 'Program on Thorium Utilization in Pressurized Water Reactors' carried out jointly by Empresas Nucleares Brasileiras S.A. (NUCLEBRAS), through its Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) and by German institutions, the Julich Nuclear Research Center (KFA), the Kraftwerk Union A.G. (KWU) and NUKEM GmbH. This paper summarizes the results so far obtained in the fuel technology. The development of a fabrication process for PWR fuel pellets from gel-microspheres is reported as well as the design, the specification, and the fabrication of prototype fuel rods for irradiation tests. (Author) [pt

  2. Molecular dynamics simulations of disjoining pressure effects in ultra-thin water films on a metal surface

    Science.gov (United States)

    Hu, Han; Sun, Ying

    2013-11-01

    Disjoining pressure, the excess pressure in an ultra-thin liquid film as a result of van der Waals interactions, is important in lubrication, wetting, flow boiling, and thin film evaporation. The classic theory of disjoining pressure is developed for simple monoatomic liquids. However, real world applications often utilize water, a polar liquid, for which fundamental understanding of disjoining pressure is lacking. In the present study, molecular dynamics (MD) simulations are used to gain insights into the effect of disjoining pressure in a water thin film. Our MD models were firstly validated against Derjaguin's experiments on gold-gold interactions across a water film and then verified against disjoining pressure in an argon thin film using the Lennard-Jones potential. Next, a water thin film adsorbed on a gold surface was simulated to examine the change of vapor pressure with film thickness. The results agree well with the classic theory of disjoining pressure, which implies that the polar nature of water molecules does not play an important role. Finally, the effects of disjoining pressure on thin film evaporation in nanoporous membrane and on bubble nucleation are discussed.

  3. 76 FR 72007 - ZionSolutions, LLC; Zion Nuclear Power Station, Units 1 and 2; Exemption From Certain Security...

    Science.gov (United States)

    2011-11-21

    ... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-295 and 50-304; NRC-2011-0244] ZionSolutions, LLC; Zion Nuclear Power Station, Units 1 and 2; Exemption From Certain Security Requirements 1.0 Background Zion Nuclear Power Station (ZNPS or Zion), Unit 1, is a Westinghouse 3250 MWt Pressurized Water Reactor...

  4. Culinary and pressure irrigation water system hydroelectric generation

    Energy Technology Data Exchange (ETDEWEB)

    Christiansen, Cory [Water Works Engineers, Pleasant Grove City, UT (United States)

    2016-01-29

    Pleasant Grove City owns and operates a drinking water system that included pressure reducing stations (PRVs) in various locations and flow conditions. Several of these station are suitable for power generation. The City evaluated their system to identify opportunities for power generation that can be implemented based on the analysis of costs and prediction of power generation and associated revenue. The evaluation led to the selection of the Battle Creek site for development of a hydro-electric power generating system. The Battle Creek site includes a pipeline that carries spring water to storage tanks. The system utilizes a PRV to reduce pressure before the water is introduced into the tanks. The evaluation recommended that the PRV at this location be replaced with a turbine for the generation of electricity. The system will be connected to the utility power grid for use in the community. A pelton turbine was selected for the site, and a turbine building and piping system were constructed to complete a fully functional power generation system. It is anticipated that the system will generate approximately 440,000 kW-hr per year resulting in $40,000 of annual revenue.

  5. Prestressed concrete pressure vessels for boiling water reactors

    International Nuclear Information System (INIS)

    Menon, S.

    1979-12-01

    Following a general description of the Scandinavian cooperative project on prestressed concrete pressure vessels for boiling water reactors, detailed discussion is given in four appendices of the following aspects: the verification programme of tests and studies, the development and testing of a liner venting system, a preliminary safety philosophy and comparative assessment of cold and hot liners. Vessel failure probability is briefly discussed and some figures presented. The pressure gradients in the vessel wall resulting from various stipulated linear cracks, with a liner venting system are presented graphically. (JIW)

  6. Water pressure control device for control rod drive

    International Nuclear Information System (INIS)

    Sato, Hideyuki.

    1981-01-01

    Purpose: To minimize the fluctuations in the reactor water level upon occurrence of abnormality by inputting the level signal of the reactor to an arithmetic unit for controlling the pressure of control rod drive water to thereby enable effective reactor level control. Constitution: Signal from a flow rate transmitter is inputted into an arithmetic unit to perform constant flow rate control upon normal operation. While on the other hand, if abnormality occurs such as feedwater pump trips, the arithmetic unit is switched from the constant flow rate control to the reactor water level control. Reactor water level signal is inputted into the arithmetic unit and the control valve is most suitably controlled, whereby water is fed from CST to the reactor by way of control rod drive water system to secure the reactor water level if feedwater to the reactor is interrupted by loss of coolants on the feedwater system. Since this enables to minimize the fluctuations in the reactor water level upon abnormality, the reactor water level can be controlled most suitably by the reactor water level signal. (Moriyama, K.)

  7. Quench pressure, thermal expulsion, and normal zone propagation in internally cooled superconductors

    International Nuclear Information System (INIS)

    Dresner, L.

    1988-01-01

    When a nonrecovering normal zone appears in an internally cooled superconductor, the pressure in the conductor rises, helium is expelled from its ends, and the normal zone grows in size. This paper presents a model of these processes that allows calculation of the pressure, the expulsion velocity, and the propagation velocity with simple formulas. The model is intended to apply to conductors such as the cable-in-conduit conductor of the Westinghouse LCT (WH-LCT) coil, the helium volumes of which have very large length-to-diameter ratios (3 /times/ 10 5 ). The predictions of the model agree with the rather limited data available from propagation experiments carried out on the WH-LCT coil. 3 refs., 1 fig

  8. 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

  9. 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

  10. Internal pressure effects in the AIRCO-LCT conductor sheath

    International Nuclear Information System (INIS)

    Luton, J.N.; Clinard, J.A.; Lue, J.W.; Gray, W.H.; Summers, L.T.; Kershaw, R.

    1985-01-01

    The large Nb 3 Sn superconducting test coil produced by Westinghouse Electric Corporation for the international Large Coil Task (LCT) utilizes a conductor composed of cabled multifilamentary strands immersed in flowing supercritical helium contained by a square structural sheath made of the high-strength stainless alloy JBX-75. Peak pressures of a few hundred atmospheres are predicted to occur during quench, and measurement of these pressures seems feasible only through penetrations of the sheath wall. Fully processed short lengths of conductor were taken from production ends, fitted with pressure taps and strain gauges, and pressurized with helium gas. Failure, at 1000 atm at liquid nitrogen temperature, was by a catastrophic splitting of the sheath at a corner. Strain measurements and burst pressure agreed with elastic-plastic finite element stress calculations made for the sheath alone. Neither the production seam weld nor the pressure tap penetrations or their fillet welds contributed to the failure, although the finite element calculations show that these areas were also highly stressed, and examination of the failed sample showed that the finite welds were of poor quality. Failure was by tensile overload, with no evidence of fatigue

  11. Startup of a high-temperature reactor cooled and moderated by supercritical-pressure light water

    International Nuclear Information System (INIS)

    Yi, Tin Tin; Ishiwatari, Yuki; Koshizuka, Seiichi; Oka, Yoshiaki

    2003-01-01

    The startup schemes of high-temperature reactors cooled and moderated by supercritical pressure light water (SCLWR-H) with square lattice and descending flow type water rods are studied by thermal-hydraulic analysis. In this study, two kinds of startup systems are investigated. In the constant pressure startup system, the reactor starts at a supercritical pressure. A flash tank and pressure reducing valves are necessary. The flash tank is designed so that the moisture content in the steam is less than 0.1%. In sliding pressure startup system, the reactor starts at a subcritical pressure. A steam-water separator and a drain tank are required for two-phase flow at startup. The separator is designed by referring to the water separator used in supercritical fossil-fired power plants. The maximum cladding surface temperature during the power-raising phase of startup is restricted not to exceed the rated value of 620degC. The minimum feedwater flow rate is 25% for constant pressure startup and 35% for sliding pressure startup system. It is found that both constant pressure startup system and sliding pressure startup system are feasible in SCLWR-H from the thermal hydraulic point of view. The core outlet temperature as high as 500degC can be achieved in the present design of SCLWR-H. Since the feedwater flow rate of SCLWR-H (1190 kg/s) is lower than that of the previous SCR designs the weight of the component required for startup is reduced. The sliding pressure startup system is better than constant pressure startup system in order to reduce the required component weight (and hence material expenditure) and to simplify the startup plant system. (author)

  12. Possibilities of tritium removal from waste waters of pressurized water reactors and fuel reprocessing plants

    International Nuclear Information System (INIS)

    Ribnikar, S.V.; Pupezin, J.D.

    1975-01-01

    Starting from parameters known for heavy water production processes, a parallel was made with separation of tritium from water. The quantity in common is the total cascade flow. The most efficient processes appear to be hydrogen sulfide, water exchange, hydrogen- and water distillation. Prospects of application of new processes are discussed briefly. Problems concerning detritiation of pressurized water reactors and large fuel reprocessing plants are analyzed. Detritiation of the former should not present problems. With the latter, economical detritiation can be achieved only after some plant flow patterns are changed. (U.S.)

  13. Durable Suit Bladder with Improved Water Permeability for Pressure and Environment Suits

    Science.gov (United States)

    Bue, Grant C.; Kuznetz, Larry; Orndoff, Evelyne; Tang, Henry; Aitchison, Lindsay; Ross, Amy

    2009-01-01

    Water vapor permeability is shown to be useful in rejecting heat and managing moisture accumulation in launch-and-entry pressure suits. Currently this is accomplished through a porous Gortex layer in the Advanced Crew and Escape Suit (ACES) and in the baseline design of the Constellation Suit System Element (CSSE) Suit 1. Non-porous dense monolithic membranes (DMM) that are available offer potential improvements for water vapor permeability with reduced gas leak. Accordingly, three different pressure bladder materials were investigated for water vapor permeability and oxygen leak: ElasthaneTM 80A (thermoplastic polyether urethane) provided from stock polymer material and two custom thermoplastic polyether urethanes. Water vapor, carbon dioxide and oxygen permeability of the DMM's was measured in a 0.13 mm thick stand-alone layer, a 0.08 mm and 0.05 mm thick layer each bonded to two different nylon and polyester woven reinforcing materials. Additional water vapor permeability and mechanical compression measurements were made with the reinforced 0.05 mm thick layers, further bonded with a polyester wicking and overlaid with moistened polyester fleece thermal underwear .This simulated the pressure from a supine crew person. The 0.05 mm thick nylon reinforced sample with polyester wicking layer was further mechanically tested for wear and abrasion. Concepts for incorporating these materials in launch/entry and Extravehicular Activity pressure suits are presented.

  14. Behavior of pressure rise and condensation caused by water evaporation under vacuum at high temperature

    International Nuclear Information System (INIS)

    Takase, Kazuyuki; Kunugi, Tomoaki; Yamazaki, Seiichiro; Fujii, Sadao

    1998-01-01

    Pressure rise and condensation characteristics during the ingress-of-coolant event (ICE) in fusion reactors were investigated using the preliminary ICE apparatus with a vacuum vessel (VV), an additional tank (AT) and an isolation valve (IV). A surface of the AT was cooled by water at RT. The high temperature and pressure water was injected into the VV which was heated up to 250degC and pressure and temperature transients in the VV were measured. The pressure increased rapidly with an injection time of the water because of the water evaporation. After the IV was opened and the VV was connected with the AT, the pressure in the VV decreased suddenly. From a series of the experiments, it was confirmed that control factors on the pressure rise were the flushing evaporation and boiling heat transfer in the VV, and then, condensation of the vapor after was effective to the depressurization in the VV. (author)

  15. Drinking water fluoride and blood pressure? An environmental study.

    Science.gov (United States)

    Amini, Hassan; Taghavi Shahri, Seyed Mahmood; Amini, Mohamad; Ramezani Mehrian, Majid; Mokhayeri, Yaser; Yunesian, Masud

    2011-12-01

    The relationship between intakes of fluoride (F) from drinking water and blood pressure has not yet been reported. We examined the relationship of F in ground water resources (GWRs) of Iran with the blood pressure of Iranian population in an ecologic study. The mean F data of the GWRs (as a surrogate for F levels in drinking water) were derived from a previously conducted study. The hypertension prevalence and the mean of systolic and diastolic blood pressures (SBP & DBP) of Iranian population by different provinces and genders were also derived from the provincial report of non-communicable disease risk factor surveillance of Iran. Statistically significant positive correlations were found between the mean concentrations of F in the GWRs and the hypertension prevalence of males (r = 0.48, p = 0.007), females (r = 0.36, p = 0.048), and overall (r = 0.495, p = 0.005). Also, statistically significant positive correlations between the mean concentrations of F in the GWRs and the mean SBP of males (r = 0.431, p = 0.018), and a borderline correlation with females (r = 0.352, p = 0.057) were found. In conclusion, we found the increase of hypertension prevalence and the SBP mean with the increase of F level in the GWRs of Iranian population.

  16. Reactor Safety Commission Code of Practice for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    1990-01-01

    The Reactor Safety Commission of the Federal German Republic has summarized in the form of Official Guidelines the safety requirements which, in the Commission's view, have to be met in the design, construction and operation of a nuclear power station equipped with a pressurized water reactor. The Third Edition of the RSK Guidelines for pressurized water reactors dated 14.10.81. is a revised and expanded version of the Second Edition dated 24.1.79. The Reactor Safety Commission will with effect from October 1981 use these Guidelines in consultations on the siting of and safety concept for the installation approval of future pressurized water reactors and will assess these nuclear power stations during their erection in the light of these Guidelines. They have not however been immediately conceived for the adaptation of existing nuclear power stations, whether under construction or in operation. The scope of application of these Guidelines to such nuclear power stations will have to be examined for each individual case. The main aim of the Guidelines is to simplify the consultation process within the reactor Safety Commission and to provide early advice on the safety requirements considered necessary by the Commission. (author)

  17. Development test procedure High Pressure Water Jet System

    International Nuclear Information System (INIS)

    Crystal, J.B.

    1995-01-01

    Development testing will be performed on the water jet cleaning fixture to determine the most effective arrangement of water jet nozzles to remove contamination from the surfaces of canisters and other debris. The following debris may be stained with dye to simulate surface contaminates: Mark O, Mark I, and Mark II Fuel Storage Canisters (both stainless steel and aluminum), pipe of various size, (steel, stainless, carbon steel and aluminum). Carbon steel and stainless steel plate, channel, angle, I-beam and other surfaces, specifically based on the Scientific Ecology Group (SEG) inventory and observations of debris within the basin. Test procedure for developmental testing of High Pressure Water Jet System

  18. Tritium issues in commercial pressurized water reactors

    International Nuclear Information System (INIS)

    Jones, G.

    2008-01-01

    Tritium has become an important radionuclide in commercial Pressurized Water Reactors because of its mobility and tendency to concentrate in plant systems as tritiated water during the recycling of reactor coolant. Small quantities of tritium are released in routine regulated effluents as liquid water and as water vapor. Tritium has become a focus of attention at commercial nuclear power plants in recent years due to inadvertent, low-level, chronic releases arising from routine maintenance operations and from component failures. Tritium has been observed in groundwater in the vicinity of stations. The nuclear industry has undertaken strong proactive corrective measures to prevent recurrence, and continues to eliminate emission sources through its singular focus on public safety and environmental stewardship. This paper will discuss: production mechanisms for tritium, transport mechanisms from the reactor through plant, systems to the environment, examples of routine effluent releases, offsite doses, basic groundwater transport and geological issues, and recent nuclear industry environmental and legal ramifications. (authors)

  19. Water-vapor pressure control in a volume

    Science.gov (United States)

    Scialdone, J. J.

    1978-01-01

    The variation with time of the partial pressure of water in a volume that has openings to the outside environment and includes vapor sources was evaluated as a function of the purging flow and its vapor content. Experimental tests to estimate the diffusion of ambient humidity through openings and to validate calculated results were included. The purging flows required to produce and maintain a certain humidity in shipping containers, storage rooms, and clean rooms can be estimated with the relationship developed here. These purging flows are necessary to prevent the contamination, degradation, and other effects of water vapor on the systems inside these volumes.

  20. Sources of radioiodine at pressurized water reactors. Final report

    International Nuclear Information System (INIS)

    Pelletier, C.A.; Cline, J.E.; Barefoot, E.D.; Hemphill, R.T.; Voilleque, P.G.; Emel, W.A.

    1978-11-01

    The report determines specific components and operations at operating pressurized water reactors that have a potential for being significant emission sources of radioactive iodine. The relative magnitudes of these specific sources in terms of the chemical forms of the radioiodine and the resultant annual averages from major components are established. The data are generalized for broad industry use for predictive purposes. The conclusions of this study indicate that the majority of radioiodine emanating from the primary side of pressurized water reactors comes from a few major areas; in some cases these sources are locally treatable; the interaction of radioiodine with plant interior surfaces is an important phenomenon mediating the source and affecting its release to the atmosphere; the chemical form varies depending on the circumstances of the release

  1. 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)

  2. Use of inexpensive pressure transducers for measuring water levels in wells

    Science.gov (United States)

    Keeland, B.D.; Dowd, J.F.; Hardegree, W.S.

    1997-01-01

    Frequent measurement of below ground water levels at multiple locations is an important component of many wetland ecosystem studies. These measurements, however, are usually time consuming, labor intensive, and expensive. This paper describes a water-level sensor that is inexpensive and easy to construct. The sensor is placed below the expected low water level in a shallow well and, when connected to a datalogger, uses a pressure transducer to detect groundwater or surface water elevations. Details of pressure transducer theory, sensor construction, calibration, and examples of field installations are presented. Although the transducers must be individually calibrated, the sensors have a linear response to changing water levels (r2 ??? .999). Measurement errors resulting from temperature fluctuations are shown to be about 4 cm over a 35??C temperature range, but are minimal when the sensors are installed in groundwater wells where temperatures are less variable. Greater accuracy may be obtained by incorporating water temperature data into the initial calibration (0.14 cm error over a 35??C temperature range). Examples of the utility of these sensors in studies of groundwater/surface water interactions and the effects of water level fluctuations on tree growth are provided. ?? 1997 Kluwer Academic Publishers.

  3. Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR [pressurized-water-reactor] plants

    International Nuclear Information System (INIS)

    Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

    1988-01-01

    Recent pressure-vessel surveillance data from the High Flux Isotope Reactor (HFIR) indicate an embrittlement fluence-rate effect that is applicable to the evaluation of the integrity of light-water reactor (LWR) pressure vessel supports. A preliminary evaluation using the HFIR data indicated increases in the nil ductility transition temperature at 32 effective full-power years (EFPY) of 100 to 130/degree/C for pressurized-water-reactor (PWR) vessel supports located in the cavity at midheight of the core. This result indicated a potential problem with regard to life expectancy. However, an accurate assessment required a detailed, specific-plant, fracture-mechanics analysis. After a survey and cursory evaluation of all LWR plants, two PWR plants that appeared to have a potential problem were selected. Results of the analyses indicate minimum critical flaw sizes small enough to be of concern before 32 EFPY. 24 refs., 16 figs., 7 tabs

  4. Theory of the Maxwell pressure tensor and the tension in a water bridge.

    Science.gov (United States)

    Widom, A; Swain, J; Silverberg, J; Sivasubramanian, S; Srivastava, Y N

    2009-07-01

    A water bridge refers to an experimental "flexible cable" made up of pure de-ionized water, which can hang across two supports maintained with a sufficiently large voltage difference. The resulting electric fields within the de-ionized water flexible cable maintain a tension that sustains the water against the downward force of gravity. A detailed calculation of the water bridge tension will be provided in terms of the Maxwell pressure tensor in a dielectric fluid medium. General properties of the dielectric liquid pressure tensor are discussed along with unusual features of dielectric fluid Bernoulli flows in an electric field. The "frictionless" Bernoulli flow is closely analogous to that of a superfluid.

  5. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    Science.gov (United States)

    McDermott, Daniel J.; Schrader, Kenneth J.; Schulz, Terry L.

    1994-01-01

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  6. RESAR-3S. Reference safety analysis report, volume 1

    International Nuclear Information System (INIS)

    1975-01-01

    The nuclear steam supply system consists of a Westinghouse pressurized water reactor and closed reactor coolant loops connected in parallel to the reactor vessel. Each loop contains a reactor coolant pump and a steam generator. Also described are an electrically heated pressurizer and certain auxiliary systems. The thermal output of the system is 3425 MW(t) and the thermal output of the core is 3411 MW(t). (FS)

  7. 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

  8. The European pressurized water reactor

    International Nuclear Information System (INIS)

    Leny, J.C.

    1993-01-01

    The present state of development of the European Pressurized Water Reactor (EPR) is outlined. During the so-called harmonization phase, the French and German utilities drew up their common requirements and evaluated the reactor concept developed until then with respect to these requirements. A main result of the harmonization phase was the issue, in September 1993, of the 'EPR Conceptual Safety Feature Review File' to be jointly assessed by the safety authorities in France and Germany. The safety objectives to be met by the EPR are specified in the second part of the paper, and some details of the primary and secondary side safety systems are given. (orig.) [de

  9. Pressure effect on the conformational equilibrium of [Leu]{sup 5}-enkephalin in water

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, A [Department of Environmental Engineering for Symbiosis, Soka University, 1-326 Tangi-cho, Hachioji, Tokyo, 192-8577 (Japan); Takekiyo, T; Yoshimura, Y [Department of Applied Chemistry, National Defence Academy, 1-10-20 Hashirimizu, Yokosuka, Kanagawa, 239-8686 (Japan); Kato, M; Taniguchi, Y, E-mail: shimizu@soka.ac.j, E-mail: take214@nda.ac.j [Department of Applied Chemistry, Ritsumeikan University, 1-1-1, Nojihigashi, Kusatsu, Shiga, 525-8577 (Japan)

    2010-03-01

    The conformational stability of [Leu]{sup 5}-enkephalin,Tyr-Gly-Gly-Phe-Leu, in water have been investigated under high pressure by FTIR spectroscopy. Three peaks at 1638, 1650, and 1680 cm{sup -1} were determined by second derivative FTIR spectra in the amide I' region of [Leu]{sup 5}-enkephalin. The peaks at 1637 and 1680 cm{sup -1} are assigned to the {beta}-strand and turn structures, respectively. These peaks mean that [Leu]{sup 5}-enkephalin takes a {beta}-hairpin-like structure in water. Moreover, the absorbance at 1638 cm{sup -1} increases with increasing pressure, and this change shows a sigmoidal curve. Thus, we concluded that [Leu]{sup 5}-enkephalin has the {beta}-hairpin-like and disordered structures in water. From the FTIR profile at high pressures, the {beta}-hairpin-like structure of [Leu]{sup 5}-enkephalin is stabilized by a high pressures. Our result shows that the folded structures such as {alpha}-helix and {beta}-hairpin structures of short peptide such as [Leu]{sup 5}-enkephalin are stabilized at high pressures.

  10. Hydrostatic pressure effect on PNIPAM cononsolvency in water-methanol solutions.

    Science.gov (United States)

    Pica, Andrea; Graziano, Giuseppe

    2017-12-01

    When methanol is added to water at room temperature and 1atm, poly (N-isopropylacrylamide), PNIPAM, undergoes a coil-to-globule collapse transition. This intriguing phenomenon is called cononsolvency. Spectroscopic measurements have shown that application of high hydrostatic pressure destroys PNIPAM cononsolvency in water-methanol solutions. We have developed a theoretical approach that identifies the decrease in solvent-excluded volume effect as the driving force of PNIPAM collapse on increasing the temperature. The same approach indicates that cononsolvency, at room temperature and P=1atm, is caused by the inability of PNIPAM to make all the attractive energetic interactions that it could be engaged in, due to competition between water and methanol molecules. The present analysis suggests that high hydrostatic pressure destroys cononsolvency because the coil state becomes more compact, and the quantity measuring PNIPAM-solvent attractions increases in magnitude due to the solution density increase, and the ability of small water molecules to substitute methanol molecules on PNIPAM surface. Copyright © 2017 Elsevier B.V. All rights reserved.

  11. Pressurized fluidized bed combustion combined cycle power plant with coal gasification: Second generation pilot plant

    International Nuclear Information System (INIS)

    Farina, G.L.; Bressan, L.

    1991-01-01

    This paper presents the technical and economical background of a research and development program of a novel power generation scheme, which is based on coal gasification, pressurized fluid bed combustion and combined cycles. The participants in this program are: Foster Wheeler (project leader), Westinghouse, IGT and the USA Dept. of Energy. The paper describes the characteristics of the plant, the research program in course of implementation, the components of the pilot plant and the first results obtained

  12. The study and improvement of water level control of pressurizer

    International Nuclear Information System (INIS)

    Gao Peng; Zhang Qinshun

    2006-01-01

    The PI controller which is used widely in water level control of pressurizer in reactor control system usually leads dynamic overshoot and long setting time. The improvement project for intelligent fuzzy controller to take the place of PI controller is advanced. This paper researches the water level control of pressurizer in reactor control system of Daya Bay Phase I, and describes the method of intelligent fuzzy control in practice. Simulation indicates that the fuzzy control has advantages of small overshoot and short settling time. It can also improve control system's real time property and anti-interference ability. Especially for non-linear and time-varying complicated control systems, it can obtain good control results. (authors)

  13. Data quality assurance in pressure transducer-based automatic water level monitoring

    Science.gov (United States)

    Submersible pressure transducers integrated with data loggers have become relatively common water-level measuring devices used in flow or well water elevation measurements. However, drift, linearity, hysteresis and other problems can lead to erroneous data. Researchers at the USDA-ARS in Watkinsvill...

  14. Comparison Of Vented And Absolute Pressure Transducers For Water-Level Monitoring In Hanford Site Central Plateau Wells

    International Nuclear Information System (INIS)

    Mcdonald, J.P.

    2011-01-01

    Automated water-level data collected using vented pressure transducers deployed in Hanford Site Central Plateau wells commonly display more variability than manual tape measurements in response to barometric pressure fluctuations. To explain this difference, it was hypothesized that vented pressure transducers installed in some wells are subject to barometric pressure effects that reduce water-level measurement accuracy. Vented pressure transducers use a vent tube, which is open to the atmosphere at land surface, to supply air pressure to the transducer housing for barometric compensation so the transducer measurements will represent only the water pressure. When using vented transducers, the assumption is made that the air pressure between land surface and the well bore is in equilibrium. By comparison, absolute pressure transducers directly measure the air pressure within the wellbore. Barometric compensation is achieved by subtracting the well bore air pressure measurement from the total pressure measured by a second transducer submerged in the water. Thus, no assumption of air pressure equilibrium is needed. In this study, water-level measurements were collected from the same Central Plateau wells using both vented and absolute pressure transducers to evaluate the different methods of barometric compensation. Manual tape measurements were also collected to evaluate the transducers. Measurements collected during this study demonstrated that the vented pressure transducers over-responded to barometric pressure fluctuations due to a pressure disequilibrium between the air within the wellbores and the atmosphere at land surface. The disequilibrium is thought to be caused by the relatively long time required for barometric pressure changes to equilibrate between land surface and the deep vadose zone and may be exacerbated by the restriction of air flow between the well bore and the atmosphere due to the presence of sample pump landing plates and well caps. The

  15. [Variation of water DOC during the process of pre-pressure and coagulation sedimentation treatment].

    Science.gov (United States)

    Chen, Wen-Jing; Cong, Hai-Bing; Xu, Ya-Jun; Wang, Wei; Jiang, Xin-Yue; Liu, Yu-Jiao

    2014-07-01

    The aim of the study was to explore whether the pre-pressure and coagulation sedimentation process would result in algal cell disruption, leading to increased dissolved organic carbon (DOC) in water, based on which, the pressure application mode would be optimized and safe and efficient pre-pressure algae removal process would be obtained. The changes in DOC during the process of pre-pressure and preoxidation treatment, the distribution of molecular weight in water as well as the removal efficiency of algae, turbidity and DOC after coagulation and sedimentation were investigated. The results showed that the DOC in water did not increase but decreased, and the molecular weight decreased after treated with 0.5-0.8 MPa pressure. While KMnO4 and NaClO pre-oxidation both increased the DOC, in the meanwhile, the distribution of molecular weight showed no obvious change. After the pre-pressure coagulation and sedimentation process, the removal rate of algae was 96.23% and that of DOC was 29. 11%, which was by 10% - 30% higher than the rate of pre-oxidation coagulation and sedimentation process.

  16. Water vapor pressure over molten KH_2PO_4 and demonstration of water electrolysis at ∼300 °C

    International Nuclear Information System (INIS)

    Berg, R.W.; Nikiforov, A.V.; Petrushina, I.M.; Bjerrum, N.J.

    2016-01-01

    Highlights: • The vapor pressure over molten KH_2PO_4 was measured by Raman spectroscopy to be about 8 bars at ∼300 °C. • Raman spectroscopy shows that molten KH_2PO_4 under its own vapor pressure contains much dissolved water. • It is demonstrated spectroscopically that water electrolysis is possible in KH_2PO_4 electrolyte forming H_2 and O_2 at 300 °C. • Molten KH_2PO_4 is a possible electrolyte for water electrolysis. - Abstract: A new potentially high-efficiency electrolyte for water electrolysis: molten monobasic potassium phosphate, KH_2PO_4 or KDP has been investigated at temperatures ∼275–325 °C. At these temperatures, KH_2PO_4 was found to dissociate into H_2O gas in equilibrium with a melt mixture of KH_2PO_4−K_2H_2P_2O_7−KPO_3−H_2O. The water vapor pressure above the melt, when contained in a closed ampoule, was determined quantitatively vs. temperature by use of Raman spectroscopy with methane or hydrogen gas as an internal calibration standard, using newly established relative ratios of Raman scattering cross sections of water and methane or hydrogen to be 0.40 ± 0.02 or 1.2 ± 0.03. At equilibrium the vapor pressure was much lower than the vapor pressure above liquid water at the same temperature. Electrolysis was realized by passing current through closed ampoules (vacuum sealed quartz glass electrolysis cells with platinum electrodes and the electrolyte melt). The formation of mixtures of hydrogen and oxygen gases as well as the water vapor was detected by Raman spectroscopy. In this way it was demonstrated that water is present in the new electrolyte: molten KH_2PO_4 can be split by electrolysis via the reaction 2H_2O → 2H_2 + O_2 at temperatures ∼275–325 °C. At these temperatures, before the start of the electrolysis, the KH_2PO_4 melt gives off H_2O gas that pressurizes the cell according to the following dissociations: 2KH_2PO_4 ↔ K_2H_2P_2O_7 + H_2O ↔ 2KPO_3 + 2H_2O. The spectra show however that the water by

  17. Intelligent Pressure Management to Reduce Leakage in Urban Water Supply Networks, A Case Study of Sarafrazan District, Mashhad

    Directory of Open Access Journals (Sweden)

    Mohammad Soltani Asl

    2009-09-01

    Full Text Available Water losses are inevitable in urban water distribution systems. The two approaches adopted nowadays to combat this problem include management of hydraulic parameters such as pressure and leakage detection in the network. Intellitgent pressure management is a suitable technique for controlling leakage and reducing damages due to high operating pressures in a network. This paper aims to investigate the effects of pressure reduction on leakage. The EPANET 2.10 software is used to simulate the water distribution network in the Sarafrazan District,Mashhad, assuming leakage from network nodes. The results are then used to develop a pressure variation program based on the patterns obtained from the simulation, which is applied to the pressure reducing valve. The results show that pressure management can reduce nightly leakage by up to 35% while maintaining a more uniform pressure distribution. Implementation of the time-dependent pressure pattern by applying programmable pressure reducing valves in a real urban water distribution network is feasible and plays a key role in reducing water losses to leakage.

  18. Toxic pressure of herbicides on microalgae in Dutch estuarine and coastal waters

    Science.gov (United States)

    Booij, Petra; Sjollema, Sascha B.; van der Geest, Harm G.; Leonards, Pim E. G.; Lamoree, Marja H.; de Voogt, W. Pim; Admiraal, Wim; Laane, Remi W. P. M.; Vethaak, A. Dick

    2015-08-01

    For several decades now, there has been an increase in the sources and types of chemicals in estuarine and coastal waters as a consequence of anthropogenic activities. This has led to considerable concern about the effects of these chemicals on the marine food chain. The fact is that estuarine and coastal waters are the most productive ecosystems with high primary production by microalgae. The toxic pressure of specific phytotoxic chemicals now poses a major threat to these ecosystems. In a previous study, six herbicides (atrazine, diuron, irgarol, isoproturon, terbutryn and terbutylazine) were identified as the main contaminants affecting photosynthesis in marine microalgae. The purpose of this study is to investigate the toxic pressure of these herbicides in the Dutch estuarine and coastal waters in relation to the effective photosystem II efficiency (ΦPSII) in microalgae. Temporal and spatial variations in the concentrations of these herbicides were analyzed based on monitoring data. Additionally, a field study was carried out in which chemical analysis of water was performed and also a toxicity assessment using the Pulse Amplitude Modulation (PAM) fluorometry assay that measures ΦPSII. The toxic pressure on ΦPSII in microalgae has decreased with 55-82% from 2003 to 2012, with the Western Scheldt estuary showing the highest toxic pressure. By combining toxicity data from the PAM assay with chemical analysis of herbicide concentrations, we have identified diuron and terbutylazine as the main contributors to the toxic pressure on microalgae. Although direct effects are not expected, the toxic pressure is close to the 10% effect level in the PAM assay. A compliance check with the current environmental legislation of the European Union revealed that the quality standards are not sufficient to protect marine microalgae.

  19. Elasticity of water-saturated rocks as a function of temperature and pressure.

    Science.gov (United States)

    Takeuchi, S.; Simmons, G.

    1973-01-01

    Compressional and shear wave velocities of water-saturated rocks were measured as a function of both pressure and temperature near the melting point of ice to confining pressure of 2 kb. The pore pressure was kept at about 1 bar before the water froze. The presence of a liquid phase (rather than ice) in microcracks of about 0.3% porosity affected the compressional wave velocity by about 5% and the shear wave velocity by about 10%. The calculated effective bulk modulus of the rocks changes rapidly over a narrow range of temperature near the melting point of ice, but the effective shear modulus changes gradually over a wider range of temperature. This phenomenon, termed elastic anomaly, is attributed to the existence of liquid on the boundary between rock and ice due to local stresses and anomalous melting of ice under pressure.

  20. Methods and means for reducing pressure in systems for fire fighting and water spraying in mines

    Energy Technology Data Exchange (ETDEWEB)

    Kozlyuk, A I; Grin' , G V; Yushchenko, Yu N

    1986-01-01

    Valves are evaluated used in water systems for fire fighting and dust suppression in underground black coal mines in the USSR. Specifications of the KR-2, the KR-3 and the R-86 pressure-reducing valves used in deep mines are analyzed. The valves are characterized by low reliability, low capacity and low pressure reducing range. Therefore groups (parallel arrangement) of pressure-reducing valves are used. Using valve groups increases equipment cost. The pressure-reducing systems should consist of no more than 2 valves. The VNIIGD Institute developed the RKGD pressure-reducing valve with the following specifications: inlet pressure 6.87 MPa, outlet pressure from 0.98 to 2.45 MPa, water discharge 100 m/sup 3//h). The RKGD valves are characterized by high reliability but extremely high weight. Therefore, the VNIIGD Institute developed a modified version of pressure-reducing valve, called the PRK (with maximum inlet pressure of 5 MPa, outlet pressure ranging from 0.5 to 1.5 MPa, water discharge 80 m/sup 3//h and weighing 5 kg). Design of the PRK pressure-reducing valve is shown.

  1. Absorber rod bundle actuator in a pressurized water nuclear reactor

    International Nuclear Information System (INIS)

    Martin, J.; Peletan, R.

    1984-01-01

    The invention concerns an absorber rod bundle actuator in a pressurized water reactor with spectral shift control. The device comprises two coaxial control bars. The inner bar is integral with the absorber rod bundle; it has an enlarged zone which acts as a proton under pressure difference across an annular seal which can be radially expanded, the pressure difference allowing to the absorber rod bundles actuating on the piston. When a pressure difference is applied, the seal expands radially by a sufficient amount to make sealing contact with the zone of larger diameter in the outer bar. The invention applies more particularly to reactors with spectral shift control using bundles of fertile rods [fr

  2. Technology transfer in the Spanish nuclear programme

    International Nuclear Information System (INIS)

    Perez-Naredo, F.

    1983-01-01

    The paper describes the process of technology transfer under the Spanish nuclear programme and its three generations of nuclear power plants during the last 20 years, with special reference to the nine new plants equipped with Westinghouse pressurized water reactors and the rising level of national involvement in these stations. It deals with the development of Westinghouse Nuclear's organization in Spain, referring to its staff and to the manufacturers who supply equipment for the programme, going into particular detail where problems of quality assurance are concerned. In conclusion, it summarizes the present capacity of Spanish industry in various areas connected with the design, manufacture and construction of nuclear power plants. (author)

  3. Pressure-induced transformations in computer simulations of glassy water

    Science.gov (United States)

    Chiu, Janet; Starr, Francis W.; Giovambattista, Nicolas

    2013-11-01

    Glassy water occurs in at least two broad categories: low-density amorphous (LDA) and high-density amorphous (HDA) solid water. We perform out-of-equilibrium molecular dynamics simulations to study the transformations of glassy water using the ST2 model. Specifically, we study the known (i) compression-induced LDA-to-HDA, (ii) decompression-induced HDA-to-LDA, and (iii) compression-induced hexagonal ice-to-HDA transformations. We study each transformation for a broad range of compression/decompression temperatures, enabling us to construct a "P-T phase diagram" for glassy water. The resulting phase diagram shows the same qualitative features reported from experiments. While many simulations have probed the liquid-state phase behavior, comparatively little work has examined the transitions of glassy water. We examine how the glass transformations relate to the (first-order) liquid-liquid phase transition previously reported for this model. Specifically, our results support the hypothesis that the liquid-liquid spinodal lines, between a low-density and high-density liquid, are extensions of the LDA-HDA transformation lines in the limit of slow compression. Extending decompression runs to negative pressures, we locate the sublimation lines for both LDA and hyperquenched glassy water (HGW), and find that HGW is relatively more stable to the vapor. Additionally, we observe spontaneous crystallization of HDA at high pressure to ice VII. Experiments have also seen crystallization of HDA, but to ice XII. Finally, we contrast the structure of LDA and HDA for the ST2 model with experiments. We find that while the radial distribution functions (RDFs) of LDA are similar to those observed in experiments, considerable differences exist between the HDA RDFs of ST2 water and experiment. The differences in HDA structure, as well as the formation of ice VII (a tetrahedral crystal), are a consequence of ST2 overemphasizing the tetrahedral character of water.

  4. Development of a quasi-adiabatic calorimeter for the determination of the water vapor pressure curve.

    Science.gov (United States)

    Mokdad, S; Georgin, E; Hermier, Y; Sparasci, F; Himbert, M

    2012-07-01

    Progress in the knowledge of the water saturation curve is required to improve the accuracy of the calibrations in humidity. In order to achieve this objective, the LNE-CETIAT and the LNE-CNAM have jointly built a facility dedicated to the measurement of the saturation vapor pressure and temperature of pure water. The principle is based on a static measurement of the pressure and the temperature of pure water in a closed, temperature-controlled thermostat, conceived like a quasi-adiabatic calorimeter. A copper cell containing pure water is placed inside a temperature-controlled copper shield, which is mounted in a vacuum-tight stainless steel vessel immersed in a thermostated bath. The temperature of the cell is measured with capsule-type standard platinum resistance thermometers, calibrated with uncertainties below the millikelvin. The vapor pressure is measured by calibrated pressure sensors connected to the cell through a pressure tube whose temperature is monitored at several points. The pressure gauges are installed in a thermostatic apparatus ensuring high stability of the pressure measurement and avoiding any condensation in the tubes. Thanks to the employment of several technical solutions, the thermal contribution to the overall uncertainty budget is reduced, and the remaining major part is mainly due to pressure measurements. This paper presents a full description of this facility and the preliminary results obtained for its characterization.

  5. STUDY OF WATER HAMMERS IN THE FILLING OF THE SYSTEM OF PRESSURE COMPENSATION IN THE WATER-COOLED AND WATER-MODERATED POWER REACTORS

    Directory of Open Access Journals (Sweden)

    A. V. Korolyev

    2017-01-01

    Full Text Available The research presented in the article conforms to the severe accident that took place at the Three Mail Island nuclear power plant in the USA. The research is focused on improving the reliability of the pressure compensator that is an important equipment of the primary circuit. In order to simulate such a situation, the stand has been developed to simulate the design of the pressurizer of the PWR-440 reactor, in particular an elliptical shape of the upper lid which has a steam outlet pipe at the top of the construction that creates conditions for occurrence of such water hammers. For the experiments, an installation has been created that makes it possible to measure and record the water hammering that occur when the tanks are filled. Measurement of the amplitude of the water hammering was carried out by a specially developed piezoelectric sensor, and the registration – by a light-beam oscilloscope. The technique of carrying out the experiment is described and the results of an experimental study of the water hammers arising when the vessels are completely filled are presented. Quantitative data were obtained on the amplitudes of the hydraulic impacts depending on the rate of filling of the vessel and the diameter of the outlet, the maximum pressure of the hydraulic shock was 7–9 atm. Comparison of calculated and experimental data has been performed. The allowable discrepancy is explained by the calculated value of the system stiffness coefficient, which did not take into account the presence of welded seams in the tank that imparts the system with additional rigidity. The calculated relationships are obtained, that make it possible to estimate the amplitudes of the water hammers through the acceleration of the water level in front of the outlet from a vessel with an elliptical bottom. The possibility of a water hammer in the pressure compensator is demonstrated by experiment and by theoretical calculations. Based on the experimental data, a

  6. Study on the pressure self-adaptive water-tight junction box in underwater vehicle

    Directory of Open Access Journals (Sweden)

    Haocai Huang

    2012-09-01

    Full Text Available Underwater vehicles play a very important role in underwater engineering. Water-tight junction box (WJB is one of the key components in underwater vehicle. This paper puts forward a pressure self-adaptive water-tight junction box (PSAWJB which improves the reliability of the WJB significantly by solving the sealing and pressure problems in conventional WJB design. By redundancy design method, the pressure self-adaptive equalizer (PSAE is designed in such a way that it consists of a piston pressure-adaptive compensator (PPAC and a titanium film pressure-adaptive compensator (TFPAC. According to hydro-mechanical simulations, the operating volume of the PSAE is more than or equal to 11.6 % of the volume of WJB liquid system. Furthermore, the required operating volume of the PSAE also increases as the gas content of oil, hydrostatic pressure or temperature difference increases. The reliability of the PSAWJB is proved by hyperbaric chamber tests.

  7. The inner containment of an EPR trademark pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ostermann, Dirk; Krumb, Christian; Wienand, Burkhard [AREVA GmbH, Offenbach (Germany)

    2014-08-15

    On February 12, 2014 the containment pressure and subsequent leak tightness tests on the containment of the Finnish Olkiluoto 3 EPR trademark reactor building were completed successfully. The containment of an EPR trademark pressurized water reactor consists of an outer containment to protect the reactor building against external hazards (such as airplane crash) and of an inner containment that is subjected to internal overpressure and high temperature in case of internal accidents. The current paper gives an overview of the containment structure, the design criteria, the validation by analyses and experiments and the containment pressure test.

  8. 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

  9. Steam condensation behavior of high pressure water's blow down directly into water in containment under LOCA

    International Nuclear Information System (INIS)

    Kusunoki, Tsuyoshi; Ishida, Toshihisa; Yoritsune, Tsutomu; Kasahara, Y.

    1995-01-01

    JAERI has been conducting a design study of an advanced type Marine Reactor X (MRX) for merchant ships. By employing 'Integral type PWR', In-vessel type control rod drive systems', 'Water filled containment system' and 'Decay heat removal system by natural convection', MRX achieved a compact, light weight and highly safe plant. Experiments on steam condensation behavior of high pressure water's blow down into water have been conducted in order to investigate a major safety issue related to the design decision of 'Water filled containment system'. (author)

  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. An improved film evaporation correlation for saline water at sub-atmospheric pressures

    KAUST Repository

    Shahzada, Muhammad Wakil; Ng, Kim Choon; Thu, Kyaw; Myat, Aung; Gee, Chun Won

    2011-01-01

    This paper presents an investigation of heat transfer correlation in a falling-film evaporator working with saline water at sub-atmospheric pressures. The experiments are conducted at different salinity levels ranging from 15000 to 90000 ppm, and the pressures were maintained between 0.92 to 2.81 kPa (corresponds to saturation temperatures of 5.9 – 23 0C). The effect of salinity, saturation pressures and chilled water temperatures on the heat transfer coefficient are accounted in the modified film evaporation correlations. The results are fitted to the Han & Fletcher's and Chun & Seban's falling-film correlations which are used in desalination industry. We modify the said correlations by adding salinity and saturation temperature corrections with respective indices to give a better agreement to our measured data.

  12. Indirect desalination of Red Sea water with forward osmosis and low pressure reverse osmosis for water reuse

    KAUST Repository

    Yangali-Quintanilla, Victor; Li, Zhenyu; Valladares Linares, Rodrigo; Li, Qingyu; Amy, Gary L.

    2011-01-01

    The use of energy still remains the main component of the costs of desalting water. Forward osmosis (FO) can help to reduce the costs of desalination, and extracting water from impaired sources can be beneficial in this regard. Experiments with FO membranes using a secondary wastewater effluent as a feed water and Red Sea water as a draw solution demonstrated that the technology is promising. FO coupled with low pressure reverse osmosis (LPRO) was implemented for indirect desalination. The system consumes only 50% (~1.5 kWh/m3) of the energy used for high pressure seawater RO (SWRO) desalination (2.5-4 kWh/m3), and produces a good quality water extracted from the impaired feed water. Fouling of the FO membranes was not a major issue during long-term experiments over 14 days. After 10 days of continuous FO operation, the initial flux declined by 28%. Cleaning the FO membranes with air scouring and clean water recovered the initial flux by 98.8%. A cost analysis revealed FO per se as viable technology. However, a minimum average FO flux of 10.5 L/m2-h is needed to compete with water reuse using UF-LPRO, and 5.5 L/m2-h is needed to recover and desalinate water at less cost than SWRO. © 2011 Elsevier B.V.

  13. Indirect desalination of Red Sea water with forward osmosis and low pressure reverse osmosis for water reuse

    KAUST Repository

    Yangali-Quintanilla, Victor

    2011-10-01

    The use of energy still remains the main component of the costs of desalting water. Forward osmosis (FO) can help to reduce the costs of desalination, and extracting water from impaired sources can be beneficial in this regard. Experiments with FO membranes using a secondary wastewater effluent as a feed water and Red Sea water as a draw solution demonstrated that the technology is promising. FO coupled with low pressure reverse osmosis (LPRO) was implemented for indirect desalination. The system consumes only 50% (~1.5 kWh/m3) of the energy used for high pressure seawater RO (SWRO) desalination (2.5-4 kWh/m3), and produces a good quality water extracted from the impaired feed water. Fouling of the FO membranes was not a major issue during long-term experiments over 14 days. After 10 days of continuous FO operation, the initial flux declined by 28%. Cleaning the FO membranes with air scouring and clean water recovered the initial flux by 98.8%. A cost analysis revealed FO per se as viable technology. However, a minimum average FO flux of 10.5 L/m2-h is needed to compete with water reuse using UF-LPRO, and 5.5 L/m2-h is needed to recover and desalinate water at less cost than SWRO. © 2011 Elsevier B.V.

  14. Initial pressure spike and its propagation phenomena in sodium-water reaction tests for MONJU steam generators

    International Nuclear Information System (INIS)

    Sato, M.; Hiroi, H.; Tanaka, N.; Hori, M.

    1977-01-01

    With the objective of demonstrating the safe design of steam generators for prototype LMFBR MONJU against the postulated large-leak accident, a number of large-leak sodium-water reaction tests have been conducted using the SWAT-1 and SWAT-3 rigs. Investigation of the potential effects of pressure load on the system is one of the major concerns in these tests. This paper reports the behavior of initial pressure spike in the reaction vessel, its propagation phenomena to the simulated secondary cooling system, and the comparisons with the computer code for one-dimensional pressure wave propagation problems. Both rigs used are the scaled-down models of the helically coiled steam generators of MONJU. The SWAT-1 rig is a simplified model and consists of a reaction vessel (1/8 scale of MONJU evaporator with 0.4 m dia. and 2.5 m height) and a pressure relief system i.e., a pressure relief line and a reaction products tank. On the other hand, the SWAT-3 rig is a 1/2.5 scale of MONJU SG system and consists of an evaporator (reaction vessel with 1.3 m dia. and 6.35 m height), a superheater, an intermediate heat exchanger (IHX), a piping system simulating the secondary cooling circuit and a pressure relief system. The both water injection systems consist of a water injection line with a rupture disk installed in front of injection hole and an electrically heated water tank. Choice of water injection rates in the scaled-down models is made based on the method of iso-velocity modeling. Test results indicated that the characteristics of the initial pressure spike are dominated by those of initial water injection which are controlled by the conditions of water heater and the size of water injection hole, etc

  15. Study on transport safety of fresh MOX fuel. Performance of the cladding tube of fresh MOX fuel against external water pressure

    International Nuclear Information System (INIS)

    Ito, Chihiro

    1999-01-01

    It is important to know the ability of the cladding tube for fresh MOX fuel against external water pressure when they were hypothetically sunk into the sea for unknown reasons. In order to evaluate the ability of cladding tubes for MOX fresh fuel against external water pressure, external water pressure tests were carried out. Resistible limit of cladding tubes against external water pressure is defined when cladding tubes are deformed largely due to buckling etc. The test results show cladding tube of BWR type can resist an external water pressure of 69 MPa (a depth of water of 7,000 m) and that of PWR type fuel can resist an external water pressure of 54 MPa (a depth of water of 5,500 m). Moreover, leak tightness is maintained at an external water pressure of 73 MPa (a depth of water of 7,400 m) for BWR type cladding tubes and at an external water pressure of 98 MPa (a depth of water of 10,000 m) for PWR type cladding tubes. (author)

  16. Limit regulation system for pressurized water nuclear reactors

    International Nuclear Information System (INIS)

    Aleite, W.; Bock, H.W.

    1976-01-01

    Described is a limit regulation system for a pressurized water nuclear reactor in combination with a steam generating system connected to a turbine, the nuclear reactor having control rods as well as an operational regulation system and a protective system, which includes reactor power limiting means operatively associated with the control rods for positioning the same and having response values between operating ranges of the operational regulation system, on the one hand, and response values of the protective system, on the other hand, and a live steam-minimal pressure regulation system cooperating with the reactor power limiting means and operatively connected to a steam inlet valve to the turbine for controlling the same

  17. Application of linearized model to the stability analysis of the pressurized water reactor

    International Nuclear Information System (INIS)

    Li Haipeng; Huang Xiaojin; Zhang Liangju

    2008-01-01

    A Linear Time-Invariant model of the Pressurized Water Reactor is formulated through the linearization of the nonlinear model. The model simulation results show that the linearized model agrees well with the nonlinear model under small perturbation. Based upon the Lyapunov's First Method, the linearized model is applied to the stability analysis of the Pressurized Water Reactor. The calculation results show that the methodology of linearization to stability analysis is conveniently feasible. (authors)

  18. Exploration of Impinging Water Spray Heat Transfer at System Pressures Near the Triple Point

    Science.gov (United States)

    Golliher, Eric L.; Yao, Shi-Chune

    2013-01-01

    The heat transfer of a water spray impinging upon a surface in a very low pressure environment is of interest to cooling of space vehicles during launch and re-entry, and to industrial processes where flash evaporation occurs. At very low pressure, the process occurs near the triple point of water, and there exists a transient multiphase transport problem of ice, water and water vapor. At the impingement location, there are three heat transfer mechanisms: evaporation, freezing and sublimation. A preliminary heat transfer model was developed to explore the interaction of these mechanisms at the surface and within the spray.

  19. Thermohydraulic analysis of pressurized water reactors

    International Nuclear Information System (INIS)

    Veloso, M.A.

    1980-01-01

    The computer program PANTERA is applied in the thermo-hydraulic analysis of Pressurized Water Reactor Cores (PWR). It is a version of COBRA-IIIC in which a new thermal conduction model for fuel rods was introduced. The results calculated by this program are compared with experimental data obtained from bundles of fuel rods, simulating reactor conditions. The validity of the new thermal model is checked too. The PANTERA code, through a simplified procedure of calculation, is used in the thermo-hydraulic analysis of Indian Point, Unit 2, reactor core, in stationary conditions. The results are discussed and compared with design data. (Autor) [pt

  20. 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