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Sample records for westinghouse two-loop pwr

  1. A reduced scale two loop PWR core designed with particle swarm optimization technique

    International Nuclear Information System (INIS)

    Lima Junior, Carlos A. Souza; Pereira, Claudio M.N.A; Lapa, Celso M.F.; Cunha, Joao J.; Alvim, Antonio C.M.

    2007-01-01

    Reduced scale experiments are often employed in engineering projects because they are much cheaper than real scale testing. Unfortunately, designing reduced scale thermal-hydraulic circuit or equipment, with the capability of reproducing, both accurately and simultaneously, all physical phenomena that occur in real scale and at operating conditions, is a difficult task. To solve this problem, advanced optimization techniques, such as Genetic Algorithms, have been applied. Following this research line, we have performed investigations, using the Particle Swarm Optimization (PSO) Technique, to design a reduced scale two loop Pressurized Water Reactor (PWR) core, considering 100% of nominal power and non accidental operating conditions. Obtained results show that the proposed methodology is a promising approach for forced flow reduced scale experiments. (author)

  2. 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.)

  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. Application of code scaling, applicability and uncertainty methodology to large break LOCA analysis of two loop PWR

    International Nuclear Information System (INIS)

    Mavko, B.; Stritar, A.; Prosek, A.

    1993-01-01

    In NED 119, No. 1 (May 1990) a series of six papers published by a Technical Program Group presented a new methodology for the safety evaluation of emergency core cooling systems in nuclear power plants. This paper describes the application of that new methodology to the LB LOCA analysis of the two loop Westinghouse power plant. Results of the original work were used wherever possible, so that the analysis was finished in less than one man year of work. Steam generator plugging level and safety injection flow rate were used as additional uncertainty parameters, which had not been used in the original work. The computer code RELAP5/MOD2 was used. Response surface was generated by the regression analysis and by the artificial neural network like Optimal Statistical Estimator method. Results were compared also to the analytical calculation. (orig.)

  5. 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

  6. Severe accident analysis in a two-loop PWR nuclear power plant with the ASTEC code

    International Nuclear Information System (INIS)

    Sadek, Sinisa; Amizic, Milan; Grgic, Davor

    2013-01-01

    The ASTEC/V2.0 computer code was used to simulate a hypothetical severe accident sequence in the nuclear power plant Krsko, a 2-loop pressurized water reactor (PWR) plant. ASTEC is an integral code jointly developed by Institut de Radioprotection et de Surete Nucleaire (IRSN, France) and Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS, Germany) to assess nuclear power plant behaviour during a severe accident. The analysis was conducted in 2 steps. First, the steady state calculation was performed in order to confirm the applicability of the plant model and to obtain correct initial conditions for the accident analysis. The second step was the calculation of the station blackout accident with a leakage of the primary coolant through degraded reactor coolant pump seals, which was a small LOCA without makeup capability. Two scenarios were analyzed: one with and one without the auxiliary feedwater (AFW). The latter scenario, without the AFW, resulted in earlier core damage. In both cases, the accident ended with a core melt and a reactor pressure vessel failure with significant release of hydrogen. In addition, results of the ASTEC calculation were compared with results of the RELAP5/SCDAPSIM calculation for the same transient scenario. The results comparison showed a good agreement between predictions of those 2 codes. (orig.)

  7. 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)

  8. 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

  9. 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

  10. 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

  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. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-15

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

  14. 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.

  15. 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

  16. Projects of Modifications of design for mitigation of accidents outside the design Bases on nuclear Central PWR Siemens-KWU and Westinghouse; Proyectos de Modificaciones de Sieno para Mitigacion de Accidentes fuera de la Bases de Diseno en Centrales Nucleares PWR Siemens-KWU y Westinghouse

    Energy Technology Data Exchange (ETDEWEB)

    Dominguez Gonzalez, G.; Cano Rodriguez, L. A.; Arguello Tara, A.

    2014-07-01

    Following the accident at the Japanese Fukushima-Daiichi NPP, the different regulators of nuclear power generation have required numerous reports regarding the evaluation and modification of the capacity of the plants to face accidents with severities beyond that established in their Design Bases. Under this new scenario, with multiple new demands and commitments, EA has carried out the required works for the implementation of strategies to mitigate the consequences of beyond Design Basis accidents for utilities owning Siemens-KWU and Westinghouse PWR nuclear power plants. (Author)

  17. 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

  18. Exxon Nuclear Company ECCS evaluation of a 2-loop Westinghouse PWR with dry containment using the ENC WREM-II ECCS model. Large break example problem

    International Nuclear Information System (INIS)

    Krajicek, J.E.

    1977-01-01

    This document is presented as a demonstration of the ENC WREM-II ECCS model calculational procedure applied to a Westinghouse 2-loop PWR with a dry containment (R. E. Ginna plant, for example). The hypothesized Loss-of-Coolant Accident (LOCA) investigated was a split break with an area equal to twice the pipe cross-sectional area. The break was assumed to occur in one pump discharge pipe (DECLS break). The analyses involved calculations using the ENC WREM-II model. The following codes were used: RELAP4-EM/ENC26A for blowdown and hot channel analyses, RELAP4-EM FLOOD/ENC26A for core reflood analysis, CONTEMPT LT/22 modified for containment backpressure analysis, and TOODEE2/APR77 for heatup analysis

  19. 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.

  20. 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)

  1. Two-Loop Splitting Amplitudes

    International Nuclear Information System (INIS)

    Bern, Z.

    2004-01-01

    Splitting amplitudes govern the behavior of scattering amplitudes at the momenta of external legs become collinear. In this talk we outline the calculation of two-loop splitting amplitudes via the unitarity sewing method. This method retains the simple factorization properties of light-cone gauge, but avoids the need for prescriptions such as the principal value or Mandelstam-Leibbrandt ones. The encountered loop momentum integrals are then evaluated using integration-by-parts and Lorentz invariance identities. We outline a variety of applications for these splitting amplitudes

  2. Two-loop splitting amplitudes

    International Nuclear Information System (INIS)

    Bern, Z.; Dixon, L.J.; Kosower, D.A.

    2004-01-01

    Splitting amplitudes govern the behavior of scattering amplitudes at the momenta of external legs become collinear. In this talk we outline the calculation of two-loop splitting amplitudes via the unitarity sewing method. This method retains the simple factorization properties of light-cone gauge, but avoids the need for prescriptions such as the principal value or Mandelstam-Leibbrandt ones. The encountered loop momentum integrals are then evaluated using integration-by-parts and Lorentz invariance identities. We outline a variety of applications for these splitting amplitudes

  3. 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.

  4. Higgs Decay to Photons at Two Loops

    International Nuclear Information System (INIS)

    Fugel, F.

    2007-01-01

    The calculation of the two-loop corrections to the partial width of an intermediate-mass Higgs boson decaying into a pair of photons is reviewed. The main focus lies on the electroweak (EW) contributions. The sum of the EW corrections ranges from -4% to 0% for a Higgs mass between 100 GeV and 150 GeV, while the complete correction at two-loop order amounts to less than ± 1.5% in this regime. (author)

  5. 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)

  6. The quark beam function at two loops

    International Nuclear Information System (INIS)

    Gaunt, Jonathan R.; Stahlhofen, Maximilian; Tackmann, Frank J.

    2014-01-01

    In differential measurements at a hadron collider, collinear initial-state radiation is described by process-independent beam functions. They are the field-theoretic analog of initial-state parton showers. Depending on the measured observable they are differential in the virtuality and/or transverse momentum of the colliding partons in addition to their usual longitudinal momentum fractions. Perturbatively, the beam functions can be calculated by matching them onto standard quark and gluon parton distribution functions. We calculate the inclusive virtuality-dependent quark beam function at NNLO, which is relevant for any observables probing the virtuality of the incoming partons, including N-jettiness and beam thrust. For such observables, our results are an important ingredient in the resummation of large logarithms at N 3 LL order, and provide all contributions enhanced by collinear t-channel singularities at NNLO for quark-initiated processes in analytic form. We perform the calculation in both Feynman and axial gauge and use two different methods to evaluate the discontinuity in the two-loop Feynman diagrams, providing nontrivial checks of the calculation. As part of our results we reproduce the known two-loop QCD splitting functions and confirm at two loops that the virtuality-dependent beam and final-state jet functions have the same anomalous dimension.

  7. Two loop integrals and QCD scattering

    International Nuclear Information System (INIS)

    Anastasiou, C.

    2001-04-01

    We present the techniques for the calculation of one- and two-loop integrals contributing to the virtual corrections to 2→2 scattering of massless particles. First, tensor integrals are related to scalar integrals with extra powers of propagators and higher dimension using the Schwinger representation. Integration By Parts and Lorentz Invariance recurrence relations reduce the number of independent scalar integrals to a set of master integrals for which their expansion in ε = 2 - D/2 is calculated using a combination of Feynman parameters, the Negative Dimension Integration Method, the Differential Equations Method, and Mellin-Barnes integral representations. The two-loop matrix-elements for light-quark scattering are calculated in Conventional Dimensional Regularisation by direct evaluation of the Feynman diagrams. The ultraviolet divergences are removed by renormalising with the MS-bar scheme. Finally, the infrared singular behavior is shown to be in agreement with the one anticipated by the application of Catani's formalism for the infrared divergences of generic QCD two-loop amplitudes. (author)

  8. Modeling of hydrogen behaviour in a PWR nuclear power plant containment with the CONTAIN code

    International Nuclear Information System (INIS)

    Bobovnik, G.; Kljenak, I.

    2001-01-01

    Hydrogen behavior in the containment during a severe accident in a two-loop Westinghouse-type PWR nuclear power plant was simulated with the CONTAIN code. The accident was initiated with a cold-leg break of the reactor coolant system in a steam generator compartment. In the input model, the containment is represented with 34 cells. Beside hydrogen concentration, the containment atmosphere temperature and pressure and the carbon monoxide concentration were observed as well. Simulations were carried out for two different scenarios: with and without successful actuation of the containment spray system. The highest hydrogen concentration occurs in the containment dome and near the hydrogen release location in the early stages of the accident. Containment sprays do not have a significant effect on hydrogen stratification.(author)

  9. Operation and maintenance in Genkai PWR Plant

    International Nuclear Information System (INIS)

    Ohta, Shojiro

    1984-01-01

    The No.1 PWR plant with 559 MW capacity in the Genkai Nuclear Power Station, Kyushu Electric Power Co., Inc., required about 115 days for the regular inspection in fiscal 1982 and thereafter, although more maintenance work was done. But No.2 plant of the same type required not more than 80 days. In most cases, the period of one operation cycle was from 10 to 12 months, but in the third operation cycle of No.2 plant, it is expected to be 13 months. The capacity ratio of the whole power station was 75.2% at the end of fiscal 1983. These operational records all exceeded the Japanese average. The plants are two-loop Westinghouse type PWRs, and No.1 plant started the commercial operation of anti h and the increment of P 0 + . (author) apacity ratio of No.1 plant was 71.6%, and that of No.2 plant was 85.5%. The intergranular attack on steam generator tubes was found first in the fifth regular inspection, and also in the sixth and seventh inspections, and the faulty tubes were plugged. The prevention of its spread is the largest problem. The in-service quality assurance activity, the personnel training program and the effort of upgrading the plant availability are reported. (Kako, I.)

  10. 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)

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

    International Nuclear Information System (INIS)

    Zhang Mingchang

    2005-01-01

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

  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. 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

  14. XLOOPS - a package calculating one- and two-loop diagrams

    International Nuclear Information System (INIS)

    Bruecher, L.

    1997-01-01

    A program package for calculating massive one- and two-loop diagrams is introduced. It consists of five parts: - a graphical user interface, - routines for generating diagrams from particle input, - procedures for calculating one-loop integrals both analytically and numerically, - routines for massive two-loop integrals, - programs for numerical integration of two-loop diagrams. Here the graphical user interface and the text interface to Maple are presented. (orig.)

  15. 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

  16. Sizewell 'B' PWR reference design

    International Nuclear Information System (INIS)

    1982-04-01

    The reference design for a PWR power station to be constructed as Sizewell 'B' is presented in 3 volumes containing 14 chapters and in a volume of drawings. The report describes the proposed design and provides the basis upon which the safety case and the Pre-Construction Safety Report have been prepared. The station is based on a 3425MWt Westinghouse PWR providing steam to two turbine generators each of 600 MW. The layout and many of the systems are based on the SNUPPS design for Callaway which has been chosen as the US reference plant for the project. (U.K.)

  17. 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

  18. 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.

  19. 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

  20. The massless two-loop two-point function

    International Nuclear Information System (INIS)

    Bierenbaum, I.; Weinzierl, S.

    2003-01-01

    We consider the massless two-loop two-point function with arbitrary powers of the propagators and derive a representation from which we can obtain the Laurent expansion to any desired order in the dimensional regularization parameter ε. As a side product, we show that in the Laurent expansion of the two-loop integral only rational numbers and multiple zeta values occur. Our method of calculation obtains the two-loop integral as a convolution product of two primitive one-loop integrals. We comment on the generalization of this product structure to higher loop integrals. (orig.)

  1. 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.

  2. Two-loop hard-thermal-loop thermodynamics with quarks

    International Nuclear Information System (INIS)

    Andersen, Jens O.; Petitgirard, Emmanuel; Strickland, Michael

    2004-01-01

    We calculate the quark contribution to the free energy of a hot quark-gluon plasma to two-loop order using hard-thermal-loop (HTL) perturbation theory. All ultraviolet divergences can be absorbed into renormalizations of the vacuum energy and the HTL quark and gluon mass parameters. The quark and gluon HTL mass parameters are determined self-consistently by a variational prescription. Combining the quark contribution with the two-loop HTL perturbation theory free energy for pure glue we obtain the total two-loop QCD free energy. Comparisons are made with lattice estimates of the free energy for N f =2 and with exact numerical results obtained in the large-N f limit

  3. Advancing PWR fuel to meet customer needs

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, F W

    1987-03-01

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

  4. Two-loop electroweak top corrections are they under control?

    CERN Document Server

    Degrassi, G.; Feruglio, F.; Gambino, P.; Vicini, A.; Degrassi, G; Fanchiotti, S; Feruglio, F; Gambino, P; Vicini, A

    1995-01-01

    The assumption that two-loop top corrections are well approximated by the O(G_mu^2 mt^4) contribution is investigated. It is shown that in the case of the ratio neutral-to-charged current amplitudes at zero momentum transfer the O(G_mu^2 mt^2 M_Z^2) terms are numerically comparable to the m_t^4 contribution for realistic values of the top mass. An estimate of the theoretical error due to unknown two-loop top effect is presented for a few observables of LEP interest.

  5. Two-loop off-shell QCD amplitudes in FDR

    CERN Document Server

    Page, Ben

    2015-01-01

    We link the FDR treatment of ultraviolet (UV) divergences to dimensional regularization up to two loops in QCD. This allows us to derive the one-loop and two-loop coupling constant and quark mass shifts necessary to translate infrared finite quantities computed in FDR to the MSbar renormalization scheme. As a by-product of our analysis, we solve a problem analogous to the breakdown of unitarity in the Four Dimensional Helicity (FDH) method beyond one loop. A fix to FDH is then presented that preserves the renormalizability properties of QCD without introducing evanescent quantities.

  6. The two-loop renormalization of general quantum field theories

    International Nuclear Information System (INIS)

    Damme, R.M.J. van.

    1984-01-01

    This thesis provides a general method to compute all first order corrections to the renormalization group equations. This requires the computation of the first perturbative corrections to the renormalization group β-functions. These corrections are described by Feynman diagrams with two loops. The two-loop renormalization is treated for an arbitrary renormalization field theory. Two cases are considered: 1. the Yukawa sector; 2. the gauge coupling and the scalar potential. In a final section, the breakdown of unitarity in the dimensional reduction scheme is discussed. (Auth.)

  7. Higgs bosons and QCD jets at two loops

    International Nuclear Information System (INIS)

    Koukoutsakis, Athanasios

    2003-04-01

    In this thesis we present techniques for the calculation of two-loop integrals contributing to the virtual corrections to physical processes with three on-shell and one off-shell external particles. First, we describe a set of basic tools that simplify the manipulation of complicated two-loop integrals. A technique for deriving helicity amplitudes with use of a set of projectors is demonstrated. Then we present an algorithm, introduced by Laporta, that helps reduce all possible two-loop integrals to a basic set of 'master integrals'. Subsequently, these master integrals are analytically evaluated by deriving and solving differential equations on the external scales of the process. Two-loop matrix elements and helicity amplitudes are calculated for the physical processes γ* → qq-barg and H → ggg respectively. Conventional Dimensional Regularization is used in the evaluation of Feynman diagrams. For both processes the infrared singular behavior is shown to agree with the one predicted by Catani. (author)

  8. Comments on two-loop Kac-Moody algebras

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, L A; Gomes, J F; Zimerman, A H [Instituto de Fisica Teorica (IFT), Sao Paulo, SP (Brazil); Schwimmer, A [Istituto Nazionale di Fisica Nucleare, Trieste (Italy)

    1991-10-01

    It is shown that the two-loop Kac-Moody algebra is equivalent to a two variable loop algebra and a decouple {beta}-{gamma} system. Similarly WZNW and CSW models having as algebraic structure the Kac-Moody algebra are equivalent to an infinity to versions of the corresponding ordinary models and decoupled Abelian fields. (author). 15 refs.

  9. Elastic ππ scattering to two loops

    International Nuclear Information System (INIS)

    Bijnens, J.; Colangelo, G.; Gasser, J.; Ecker, G.; Sainio, M.E.

    1995-11-01

    We evaluate analytically the elastic ππ scattering amplitude to two loops in chiral perturbation theory and give numerical values for the two S-wave scattering lengths and for the phase shift difference δ 0 0 -δ 1 1 . (author)

  10. Heavy quark form factors at two loops in perturbative QCD

    International Nuclear Information System (INIS)

    Ablinger, J.; Schneider, C.; Behring, A.; Falcioni, G.

    2017-11-01

    We present the results for heavy quark form factors at two-loop order in perturbative QCD for different currents, namely vector, axial-vector, scalar and pseudo-scalar currents, up to second order in the dimensional regularization parameter. We outline the necessary computational details, ultraviolet renormalization and corresponding universal infrared structure.

  11. The two loop superstring vacuum amplitude and canonical divisors

    International Nuclear Information System (INIS)

    Parkes, A.

    1989-01-01

    I use the prescription of placing the picture changing operators at the zeroes of some holomorphic one-form and calculate the two loop superstring vacuum amplitude in the language of theta functions. It vanishes pointwise on moduli space after the use of Fay's trisecant identity and generalised Riemann identities. I briefly discuss the higher genus case. (orig.)

  12. Two-loop matching coefficients for heavy quark currents

    International Nuclear Information System (INIS)

    Kniehl, B.A.; Onishchenko, A.; Petersburg Nuclear Physics Institute, Gatchina; Piclum, J.H.; Karlsruhe Univ.; Steinhauser, M.

    2006-04-01

    In this paper we consider the matching coefficients up to two loops between Quantum Chromodynamics (QCD) and Non-Relativistic QCD (NRQCD) for the vector, axial-vector, scalar and pseudo-scalar currents. The structure of the effective theory is discussed and analytical results are presented. Particular emphasis is put on the singlet diagrams. (Orig.)

  13. Two-loop neutrino model with exotic leptons

    Science.gov (United States)

    Okada, Hiroshi; Orikasa, Yuta

    2016-01-01

    We propose a two-loop induced neutrino mass model, in which we show some bench mark points to satisfy the observed neutrino oscillation, the constraints of lepton flavor violations, and the relic density in the coannihilation system satisfying the current upper bound on the spin independent scattering cross section with nuclei. We also discuss new sources of muon anomalous magnetic moments.

  14. Two-loop polygon Wilson loops in N = 4 SYM

    International Nuclear Information System (INIS)

    Anastasiou, C.; Brandhuber, A.; Heslop, P.; Spence, B.; Travaglini, G.; Khoze, V.V.

    2009-01-01

    We compute for the first time the two-loop corrections to arbitrary n-gon lightlike Wilson loops in N = 4 supersymmetric Yang-Mills theory, using efficient numerical methods. The calculation is motivated by the remarkable agreement between the finite part of planar six-point MHV amplitudes and hexagon Wilson loops which has been observed at two loops. At n = 6 we confirm that the ABDK/BDS ansatz must be corrected by adding a remainder function, which depends only on conformally invariant ratios of kinematic variables. We numerically compute remainder functions for n = 7,8 and verify dual conformal invariance. Furthermore, we study simple and multiple collinear limits of the Wilson loop remainder functions and demonstrate that they have precisely the form required by the collinear factorisation of the corresponding two-loop n-point amplitudes. The number of distinct diagram topologies contributing to the n-gon Wilson loops does not increase with n, and there is a fixed number of 'master integrals', which we have computed. Thus we have essentially computed general polygon Wilson loops, and if the correspondence with amplitudes continues to hold, all planar n-point two-loop MHV amplitudes in the N = 4 theory.

  15. The heavy quark form factors at two loops

    Energy Technology Data Exchange (ETDEWEB)

    Ablinger, J.; Schneider, C. [Johannes Kepler Univ., Linz (Austria). Research Inst. for Symbolic Computation (RISC); Behring, A. [Deutsches Elektronen-Synchrotron (DESY), Zeuthen (Germany); RWTH Aachen Univ. (Germany). Inst. fuer Theoretische Teilchenphysik und Kosmologie; Bluemlein, J.; Freitas, A. de; Marquard, P.; Rana, N. [Deutsches Elektronen-Synchrotron (DESY), Zeuthen (Germany); Falcioni, G. [Deutsches Elektronen-Synchrotron (DESY), Zeuthen (Germany); Nikhef, Amsterdam (Netherlands). Theory Group

    2017-12-15

    We compute the two-loop QCD corrections to the heavy quark form factors in case of the vector, axial-vector, scalar and pseudo-scalar currents up to second order in the dimensional parameter ε=(4-D)/2. These terms are required in the renormalization of the higher order corrections to these form factors.

  16. Supersymmetric Regularization Two-Loop QCD Amplitudes and Coupling Shifts

    International Nuclear Information System (INIS)

    Dixon, Lance

    2002-01-01

    We present a definition of the four-dimensional helicity (FDH) regularization scheme valid for two or more loops. This scheme was previously defined and utilized at one loop. It amounts to a variation on the standard 't Hooft-Veltman scheme and is designed to be compatible with the use of helicity states for ''observed'' particles. It is similar to dimensional reduction in that it maintains an equal number of bosonic and fermionic states, as required for preserving supersymmetry. Supersymmetry Ward identities relate different helicity amplitudes in supersymmetric theories. As a check that the FDH scheme preserves supersymmetry, at least through two loops, we explicitly verify a number of these identities for gluon-gluon scattering (gg → gg) in supersymmetric QCD. These results also cross-check recent non-trivial two-loop calculations in ordinary QCD. Finally, we compute the two-loop shift between the FDH coupling and the standard MS coupling, α s . The FDH shift is identical to the one for dimensional reduction. The two-loop coupling shifts are then used to obtain the three-loop QCD β function in the FDH and dimensional reduction schemes

  17. Supersymmetric Regularization Two-Loop QCD Amplitudes and Coupling Shifts

    Energy Technology Data Exchange (ETDEWEB)

    Dixon, Lance

    2002-03-08

    We present a definition of the four-dimensional helicity (FDH) regularization scheme valid for two or more loops. This scheme was previously defined and utilized at one loop. It amounts to a variation on the standard 't Hooft-Veltman scheme and is designed to be compatible with the use of helicity states for ''observed'' particles. It is similar to dimensional reduction in that it maintains an equal number of bosonic and fermionic states, as required for preserving supersymmetry. Supersymmetry Ward identities relate different helicity amplitudes in supersymmetric theories. As a check that the FDH scheme preserves supersymmetry, at least through two loops, we explicitly verify a number of these identities for gluon-gluon scattering (gg {yields} gg) in supersymmetric QCD. These results also cross-check recent non-trivial two-loop calculations in ordinary QCD. Finally, we compute the two-loop shift between the FDH coupling and the standard {bar M}{bar S} coupling, {alpha}{sub s}. The FDH shift is identical to the one for dimensional reduction. The two-loop coupling shifts are then used to obtain the three-loop QCD {beta} function in the FDH and dimensional reduction schemes.

  18. Finite volume at two-loops in chiral perturbation theory

    International Nuclear Information System (INIS)

    Bijnens, Johan; Rössler, Thomas

    2015-01-01

    We calculate the finite volume corrections to meson masses and decay constants in two and three flavour Chiral Perturbation Theory to two-loop order. The analytical results are compared with the existing result for the pion mass in two-flavour ChPT and the partial results for the other quantities. We present numerical results for all quantities.

  19. Two-loop feed water control system in BWR plants

    International Nuclear Information System (INIS)

    Omori, Takashi; Watanabe, Takao; Hirose, Masao.

    1982-01-01

    In the process of the start-up and shutdown of BWR plants, the operation of changing over feed pumps corresponding to plant output is performed. Therefore, it is necessary to develop the automatic changeover system for feed pumps, which minimizes the variation of water level in reactors and is easy to operate. The three-element control system with the water level in reactors, the flow rate of main steam and the flow rate of feed water as the input is mainly applied, but long time is required for the changeover of feed pumps. The two-loop feed control system can control simultaneously two pumps being changed over, therefore it is suitable to the automatic changeover control system for feed pumps. Also it is excellent for the control of the recirculating valves of feed pumps. The control characteristics of the two-loop feed water control system against the external disturbance which causes the variation of water level in reactors were examined. The results of analysis by simulation are reported. The features of the two-loop feed water control system, the method of simulation and the evaluation of the two-loop feed water control system are described. Its connection with a digital feed water recirculation control system is expected. (Kako, I.)

  20. Heavy quark form factors at two loops in perturbative QCD

    Energy Technology Data Exchange (ETDEWEB)

    Ablinger, J.; Schneider, C. [Johannes Kepler Univ., Linz (Austria). Research Inst. for Symbolic Computation (RISC); Behring, A. [RWTH Aachen Univ. (Germany). Inst. fuer Theoretische Teilchenphysik und Kosmologie; Bluemlein, J.; Freitas, A. de; Marquard, P.; Rana, N. [Deutsches Elektronen-Synchrotron (DESY), Zeuthen (Germany); Falcioni, G. [Nikhef, Amsterdam (Netherlands). Theory Group

    2017-11-15

    We present the results for heavy quark form factors at two-loop order in perturbative QCD for different currents, namely vector, axial-vector, scalar and pseudo-scalar currents, up to second order in the dimensional regularization parameter. We outline the necessary computational details, ultraviolet renormalization and corresponding universal infrared structure.

  1. New methodology for the analysis of the quality controls of ENUSA on PWR components; Nueva metodologia para el analisis de los controles de calidad de ENUSA sobre los componentes PWR

    Energy Technology Data Exchange (ETDEWEB)

    Munoz, A.; Navas, I. de; Prieto, M.

    2012-07-01

    For the manufacture of PWR fuel assemblies, ENUSA receives the components of Westinghouse, who ensures its quality. However, ENUSA carried out on these components various quality controls that increase reliability and give added value.

  2. 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)

  3. Two-loop SL(2) form factors and maximal transcendentality

    International Nuclear Information System (INIS)

    Loebbert, Florian; Sieg, Christoph; Wilhelm, Matthias; Yang, Gang

    2016-01-01

    Form factors of composite operators in the SL(2) sector of N=4 SYM theory are studied up to two loops via the on-shell unitarity method. The non-compactness of this subsector implies the novel feature and technical challenge of an unlimited number of loop momenta in the integrand’s numerator. At one loop, we derive the full minimal form factor to all orders in the dimensional regularisation parameter. At two loops, we construct the complete integrand for composite operators with an arbitrary number of covariant derivatives, and we obtain the remainder functions as well as the dilatation operator for composite operators with up to three covariant derivatives. The remainder functions reveal curious patterns suggesting a hidden maximal uniform transcendentality for the full form factor. Finally, we speculate about an extension of these patterns to QCD.

  4. Two-loop SL(2) form factors and maximal transcendentality

    Energy Technology Data Exchange (ETDEWEB)

    Loebbert, Florian [Institut für Physik, Humboldt-Universität zu Berlin,Zum Großen Windkanal 6, 12489 Berlin (Germany); Sieg, Christoph [Institut für Physik, Humboldt-Universität zu Berlin,Zum Großen Windkanal 6, 12489 Berlin (Germany); Institut für Mathematik, Humboldt-Universität zu Berlin,Zum Großen Windkanal 6, 12489 Berlin (Germany); Wilhelm, Matthias [Institut für Physik, Humboldt-Universität zu Berlin,Zum Großen Windkanal 6, 12489 Berlin (Germany); Institut für Mathematik, Humboldt-Universität zu Berlin,Zum Großen Windkanal 6, 12489 Berlin (Germany); Niels Bohr Institute, Copenhagen University,Blegdamsvej 17, 2100 Copenhagen Ø (Denmark); Yang, Gang [CAS Key Laboratory of Theoretical Physics,Institute of Theoretical Physics, Chinese Academy of Sciences,Beijing 100190 (China); Institut für Physik, Humboldt-Universität zu Berlin,Zum Großen Windkanal 6, 12489 Berlin (Germany)

    2016-12-19

    Form factors of composite operators in the SL(2) sector of N=4 SYM theory are studied up to two loops via the on-shell unitarity method. The non-compactness of this subsector implies the novel feature and technical challenge of an unlimited number of loop momenta in the integrand’s numerator. At one loop, we derive the full minimal form factor to all orders in the dimensional regularisation parameter. At two loops, we construct the complete integrand for composite operators with an arbitrary number of covariant derivatives, and we obtain the remainder functions as well as the dilatation operator for composite operators with up to three covariant derivatives. The remainder functions reveal curious patterns suggesting a hidden maximal uniform transcendentality for the full form factor. Finally, we speculate about an extension of these patterns to QCD.

  5. Two-loop renormalization of quantum gravity simplified

    Science.gov (United States)

    Bern, Zvi; Chi, Huan-Hang; Dixon, Lance; Edison, Alex

    2017-02-01

    The coefficient of the dimensionally regularized two-loop R3 divergence of (nonsupersymmetric) gravity theories has recently been shown to change when nondynamical three-forms are added to the theory, or when a pseudoscalar is replaced by the antisymmetric two-form field to which it is dual. This phenomenon involves evanescent operators, whose matrix elements vanish in four dimensions, including the Gauss-Bonnet operator which is also connected to the trace anomaly. On the other hand, these effects appear to have no physical consequences for renormalized scattering processes. In particular, the dependence of the two-loop four-graviton scattering amplitude on the renormalization scale is simple. We explain this result for any minimally-coupled massless gravity theory with renormalizable matter interactions by using unitarity cuts in four dimensions and never invoking evanescent operators.

  6. Two-loop Dirac neutrino mass and WIMP dark matter

    OpenAIRE

    Bonilla, Cesar; Ma, Ernest; Peinado, Eduardo; Valle, Jose W.F.

    2018-01-01

    We propose a "scotogenic" mechanism relating small neutrino mass and cosmological dark matter. Neutrinos are Dirac fermions with masses arising only in two--loop order through the sector responsible for dark matter. Two triality symmetries ensure both dark matter stability and strict lepton number conservation at higher orders. A global spontaneously broken U(1) symmetry leads to a physical $Diracon$ that induces invisible Higgs decays which add up to the Higgs to dark matter mode. This enhan...

  7. 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.

  8. 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

  9. 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

  10. 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

  11. Parton-parton scattering at two-loops

    International Nuclear Information System (INIS)

    Tejeda Yeomans, M.E.

    2001-01-01

    Abstract We present an algorithm for the calculation of scalar and tensor one- and two-loop integrals that contribute to the virtual corrections of 2 → 2 partonic scattering. First, the tensor integrals are related to scalar integrals that contain an irreducible propagator-like structure in the numerator. Then, we use Integration by Parts and Lorentz Invariance recurrence relations to build a general system of equations that enables the reduction of any scalar integral (with and without structure in the numerator) to a basis set of master integrals. Their expansions in ε = 2 - D/2 have already been calculated and we present a summary of the techniques that have been used to this end, as well as a compilation of the expansions we need in the different physical regions. We then apply this algorithm to the direct evaluation of the Feynman diagrams contributing to the O(α s 4 ) one- and two-loop matrix-elements for massless like and unlike quark-quark, quark-gluon and gluon-gluon scattering. The analytic expressions we provide are regularised in Convensional Dimensional Regularisation and renormalised in the MS-bar scheme. Finally, we show that the structure of the infrared divergences agrees with that predicted by the application of Catani's formalism to the analysis of each partonic scattering process. The results presented in this thesis provide the complete calculation of the one- and two-loop matrix-elements for 2 → 2 processes needed for the next-to-next-to-leading order contribution to inclusive jet production at hadron colliders. (author)

  12. Two-loop ladder diagram contributions to Bhabha scattering. III

    International Nuclear Information System (INIS)

    Bjoerkevoll, K.S.; Osland, P.; Faeldt, G.

    1992-01-01

    The authors evaluate, in the high-energy limit, the sum of the Feynman amplitudes corresponding the six two-loop ladder-like diagrams in Bhabha scattering. This is the limit where s→∞, while t, the electron mass m and the photon mass λ are all being held fixed. In this limit the sum of the six Feynman amplitudes does not depend on the electron mass. When specialized to the region s>>t>>m 2 >>λ 2 , this result complements the one previously obtained. The connection with Φ 3 theory is also investigated. 6 refs

  13. Two-loop string theory on null compactifications

    International Nuclear Information System (INIS)

    Cove, Henry C.D.; Szabo, Richard J.

    2006-01-01

    We compute the two-loop contributions to the free energy in the null compactification of perturbative string theory at finite temperature. The cases of bosonic, type II and heterotic strings are all treated. The calculation exploits an explicit reductive parametrization of the moduli space of infinite-momentum frame string worldsheets in terms of branched cover instantons. Various arithmetic and physical properties of the instanton sums are described. Applications to symmetric product orbifold conformal field theories and to the matrix string theory conjecture are also briefly discussed

  14. Status of developing advanced PWR in Japan

    International Nuclear Information System (INIS)

    Iida, Yotaro

    1982-01-01

    During past eleven years since the first PWR power plant, Mihama Unit 1 of Kansai Electric Power Co., started the commercial operation in 1970, Mitsubishi Heavy Industries has endeavored to improve PWR technologies on the basis of the advice from electric power companies and the technical information to overcome difficulties in PWR power plants. Now, the main objective is to improve the overall plant performance, and the rate of operation of Japanese PWR power plants has significantly risen. The improvement of the reliability, the shortening of regular inspection period and the reduction of radioactive waste handling were attempted. In view of the satisfactory operational experience of Westinghouse type PWRs, the basic reactor concept has not been changed so far. Mitsubishi and Westinghouse reached basic agreement in August, 1981, to develop a spectral shift type large capacity reactor as the advanced PWRs for Japan. This type of PWRs hab higher degree of freedom for extended fuel cycle operation and enhances the advantage of entire fuel cycle economy, particularly the significant reduction of uranium use. The improved neutron economy is attainable by reducing neutron loss, and the core design with low power density and the economical use of plutonium are advantageous for the fuel cycle economy. (Kako, I.)

  15. Two-Loop Scattering Amplitudes from the Riemann Sphere

    CERN Document Server

    Geyer, Yvonne; Monteiro, Ricardo; Tourkine, Piotr

    2016-01-01

    The scattering equations give striking formulae for massless scattering amplitudes at tree level and, as shown recently, at one loop. The progress at loop level was based on ambitwistor string theory, which naturally yields the scattering equations. We proposed that, for ambitwistor strings, the standard loop expansion in terms of the genus of the worldsheet is equivalent to an expansion in terms of nodes of a Riemann sphere, with the nodes carrying the loop momenta. In this paper, we show how to obtain two-loop scattering equations with the correct factorization properties. We adapt genus-two integrands from the ambitwistor string to the nodal Riemann sphere and show that these yield correct answers, by matching standard results for the four-point two-loop amplitudes of maximal supergravity and super-Yang-Mills theory. In the Yang-Mills case, this requires the loop analogue of the Parke-Taylor factor carrying the colour dependence, which includes non-planar contributions.

  16. 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)

  17. 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)

  18. 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)

  19. Two-loop fermionic corrections to massive Bhabha scattering

    Energy Technology Data Exchange (ETDEWEB)

    Actis, S.; Riemann, T. [Deutsches Elektronen-Synchrotron (DESY), Zeuthen (Germany); Czakon, M. [Wuerzburg Univ. (Germany). Inst. fuer Theoretische Physik und Astrophysik]|[Institute of Nuclear Physics, NSCR DEMOKRITOS, Athens (Greece); Gluza, J. [Silesia Univ., Katowice (Poland). Inst. of Physics

    2007-05-15

    We evaluate the two-loop corrections to Bhabha scattering from fermion loops in the context of pure Quantum Electrodynamics. The differential cross section is expressed by a small number of Master Integrals with exact dependence on the fermion masses m{sub e}, m{sub f} and the Mandelstam invariants s, t, u. We determine the limit of fixed scattering angle and high energy, assuming the hierarchy of scales m{sup 2}{sub e}<

  20. Massive two-loop Bhabha scattering - the factorizable subset

    International Nuclear Information System (INIS)

    Fleischer, J.; Tarasov, O.V.; Werthenbach, A.

    2002-11-01

    The experimental precision that will be reached at the next generation of colliders makes it indispensable to improve theoretical predictions significantly. Bhabha scattering (e + e - → e + e - ) is one of the prime processes calling for a better theoretical precision, in particular for non-zero electron masses. We present a first subset of the full two-loop calculation, namely the factorizable subset. Our calculation is based on DIANA. We reduce tensor integrals to scalar integrals in shifted (increased) dimensions and additional powers of various propagators, so-called dots-on-lines. Recurrence relations remove those dots-on-lines as well as genuine dots-on-lines (originating from mass renormalization) and reduce the dimension of the integrals to the generic d=4-2ε dimensions. The resulting master integrals have to be expanded to O(ε) to ensure proper treatment of all finite terms. (orig.)

  1. A two-loop test of M(atrix) theory

    International Nuclear Information System (INIS)

    Becker, K.

    1997-01-01

    We consider the scattering of two Dirichlet zero-branes in M(atrix) theory. Using the formulation of M(atrix) theory in terms of ten-dimensional super Yang-Mills theory dimensionally reduced to (0+1) dimensions, we obtain the effective (velocity-dependent) potential describing these particles. At one loop we obtain the well-known result for the leading order of the effective potential V eff ∝v 4 /r 7 , where v and r are the relative velocity and distance between the two zero-branes, respectively. A calculation of the effective potential at two loops shows that no renormalizations of the v 4 term of the effective potential occur at this order. (orig.)

  2. 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

  3. 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

  4. 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

  5. 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)

  6. French PWR Safety Philosophy

    International Nuclear Information System (INIS)

    Conte, M. M.

    1986-01-01

    The first 900 MWe units, built under the American Westinghouse licence and with reference to the U. S. regulation, were followed by 28 standardized units, C P1 and C P2 series. Increasing knowledge and lessons learned from starting and operating experience of French nuclear power plants, completed by the experience learned from the operation of foreign reactors, has contributed to the improvement of French PWR design and safety philosophy. As early as 1976, this experience was taken into account by French Safety organisms to discuss, with Electricite de France, the safety options for the planned 1300 MWe units, P4 and P4 series. In 1983, the new reactor scheduled, Ni4 series 1400 MWe, is a totally French design which satisfies the French regulations and other French standards and codes. Based on a deterministic approach, the French safety analysis was progressively completed by a probabilistic approach each of them having possibilities and limits. Increasing knowledge and lessons learned from operating experience have contributed to the French safety philosophy improvement. The methodology now applied to safety evaluation develops a new facet of the in depth defense concept by taking highly unlikely events into consideration, by developing the search of safety consistency of the design, and by completing the deterministic approach by the probabilistic one

  7. The two-loop sunrise integral and elliptic polylogarithms

    Energy Technology Data Exchange (ETDEWEB)

    Adams, Luise; Weinzierl, Stefan [Institut fuer Physik, Johannes Gutenberg-Universitaet Mainz (Germany); Bogner, Christian [Institut fuer Physik, Humboldt-Universitaet zu Berlin (Germany)

    2016-07-01

    In this talk, we present a solution for the two-loop sunrise integral with arbitrary masses around two and four space-time dimensions in terms of a generalised elliptic version of the multiple polylogarithms. Furthermore we investigate the elliptic polylogarithms appearing in higher orders in the dimensional regularisation ε of the two-dimensional equal mass solution. Around two space-time dimensions the solution consists of a sum of three elliptic dilogarithms where the arguments have a nice geometric interpretation as intersection points of the integration region and an elliptic curve associated to the sunrise integral. Around four space-time dimensions the sunrise integral can be expressed with the ε{sup 0}- and ε{sup 1}-solution around two dimensions, mass derivatives thereof and simpler terms. Considering higher orders of the two-dimensional equal mass solution we find certain generalisations of the elliptic polylogarithms appearing in the ε{sup 0}- and ε{sup 1}-solutions around two and four space-time dimensions. We show that these higher order-solutions can be found by iterative integration within this class of functions.

  8. 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

  9. BEACON TSM application system to the operation of PWR reactors

    International Nuclear Information System (INIS)

    Lozano, J. A.; Mildrum, C.; Serrano, J. F.

    2012-01-01

    BEACON-TSM is an advanced core monitoring system for PWR reactor cores, and also offers the possibility to perform a wide range of predictive calculation in support of reactor operation. BEACON-TSM is presently installed and licensed in the 5 Spanish PWR reactors of standard Westinghouse design. the purpose of this paper is to describe the features of this software system and to show the advantages obtainable by a nuclear power plant from its use. To illustrate the capabilities and benefits of BEACON-TSM two real case reactor operating situations are presented. (Author)

  10. Renormalization of vacuum expectation values in spontaneously broken gauge theories: two-loop results

    International Nuclear Information System (INIS)

    Sperling, Marcus; Stöckinger, Dominik; Voigt, Alexander

    2014-01-01

    We complete the two-loop calculation of β-functions for vacuum expectation values (VEVs) in gauge theories by the missing O(g 4 )-terms. The full two-loop results are presented for generic and supersymmetric theories up to two-loop level in arbitrary R ξ -gauge. The results are obtained by means of a scalar background field, identical to our previous analysis. As a by-product, the two-loop scalar anomalous dimension for generic supersymmetric theories is presented. As an application we compute the β-functions for VEVs and tan β in the MSSM, NMSSM, and E 6 SSM

  11. Uranium savings on a once through PWR fuel cycle

    International Nuclear Information System (INIS)

    Cupo, J.V.

    1980-01-01

    A number of alternatives which have the greatest potential for near term savings with minimum plant and fuel modifications have been examined at Westinghouse as part of continued internal assessment and part of NASAP study conducted for DOE pertaining to uranium utilization in a once through PWR fuel cycle. The alternatives which could be retrofitted to existing reactors were examined in more detail in the evaluation since they would have the greater near term impact on U savings

  12. 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)

  13. Semi-automatic ultrasonic inspection of PWR upper internal immersed components

    International Nuclear Information System (INIS)

    Dombret, P.; Coquette, A.; Cermak, J.; Verspeelt, D.

    1985-01-01

    The present paper describes the characteristics of a semi-automatic ultrasonic inspection system. Components inspected are the so-called flexures, small pins located at the upper part of control rod tube-guide, some of which happened to broke in a few Westinghouse type PWR's. Inspection results and other system capabilities are also mentioned

  14. PHEDRE model for the simulation of PWR reactors

    International Nuclear Information System (INIS)

    Bernard, Patrice; Dupraz, Remy; Vasile, Alfredo.

    1979-11-01

    This note presents the model of PHEDRE, simulator of a PWR, set on the hybrid computers of CISI, at the Nuclear Research Center of Cadarache. The model mainly concerns the primary part and the steam production of the PWR constructed in France. It includes an axial modelization of the core, the pressurizer, two loops of steam production and the inlet of the turbine, and the regulations concerning these components. The note presents the equations of the model, the structures of the codes concerning the initialization and the dynamic resolution, and describes the control panel of PHEDRE [fr

  15. ROX PWR

    International Nuclear Information System (INIS)

    Akie, H.; Yamashita, T.; Shirasu, N.; Takano, H.; Anoda, Y.; Kimura, H.

    1999-01-01

    For an efficient burnup of excess plutonium from nuclear reactors spent fuels and dismantled warheads, plutonium rock-like oxide(ROX) fuel has been investigated. The ROX fuel is expected to provide high Pu transmutation capability, irradiation stability and chemical and geological stability. While, a zirconia-based ROX(Zr-ROX)-fueled PWR core has some problems of Doppler reactivity coefficient and power peaking factor. For the improvement of these characteristics, two approaches were considered: the additives such as UO 2 , ThO 2 and Er 2 O 3 , and a heterogeneous core with Zr-ROX and UO 2 assemblies. As a result, the additives UO 2 + Er 2 O 3 are found to sufficiently improve the reactivity coefficients and accident behavior, and to flatten power distribution. On the other hand, in the 1/3Zr-ROX + 2/3UO 2 heterogeneous core, further reduction of power peaking seems necessary. (author)

  16. 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)

  17. 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

  18. 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.)

  19. MS vs. pole masses of gauge bosons II: Two-loop electroweak fermion correct

    International Nuclear Information System (INIS)

    Jegerlehner, F.; Kalmykov, M.Yu.; Veretin, O.

    2002-12-01

    We have calculated the fermion contributions to the shift of the position of the poles of the massive gauge boson propagators at two-loop order in the Standard Model. Together with the bosonic contributions calculated previously the full two-loop corrections are available. This allows us to investigate the full correction in the relationship between anti M anti S and pole masses of the vector bosons Z and W. Two-loop renormalization and the corresponding renormalization group equations are discussed. Analytical results for the master-integrals appearing in the massless fermion contributions are given. A new approach of summing multiple binomial sums has been developed. (orig.)

  20. Iterative structure within the five-particle two-loop amplitude

    International Nuclear Information System (INIS)

    Cachazo, Freddy; Spradlin, Marcus; Volovich, Anastasia

    2006-01-01

    We find an unexpected iterative structure within the two-loop five-gluon amplitude in N=4 supersymmetric Yang-Mills theory. Specifically, we show that a subset of diagrams contributing to the full amplitude, including a two-loop pentagon-box integral with nontrivial dependence on five kinematical variables, satisfies an iterative relation in terms of one-loop scalar box diagrams. The implications of this result for the possible iterative structure of the full two-loop amplitude are discussed

  1. ROX PWR

    Energy Technology Data Exchange (ETDEWEB)

    Akie, H.; Yamashita, T.; Shirasu, N.; Takano, H.; Anoda, Y.; Kimura, H. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    For an efficient burnup of excess plutonium from nuclear reactors spent fuels and dismantled warheads, plutonium rock-like oxide(ROX) fuel has been investigated. The ROX fuel is expected to provide high Pu transmutation capability, irradiation stability and chemical and geological stability. While, a zirconia-based ROX(Zr-ROX)-fueled PWR core has some problems of Doppler reactivity coefficient and power peaking factor. For the improvement of these characteristics, two approaches were considered: the additives such as UO{sub 2}, ThO{sub 2} and Er{sub 2}O{sub 3}, and a heterogeneous core with Zr-ROX and UO{sub 2} assemblies. As a result, the additives UO{sub 2}+ Er{sub 2}O{sub 3} are found to sufficiently improve the reactivity coefficients and accident behavior, and to flatten power distribution. On the other hand, in the 1/3Zr-ROX + 2/3UO{sub 2} heterogeneous core, further reduction of power peaking seems necessary. (author)

  2. 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.

  3. 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

  4. Large momentum expansion of two-loop self-energy diagrams with arbitrary masses

    International Nuclear Information System (INIS)

    Davydychev, A.I.; Smirnov, V.A.; Tausk, J.B.

    1993-01-01

    For two-loop two-point diagrams with arbitrary masses, an algorithm to derive the asymptotic expansion at large external momentum squared is constructed. By using a general theorem on asymptotic expansions of Feynman diagrams, the coefficients of the expansion are calculated analytically. For some two-loop diagrams occurring in the Standard Model, comparison with results of numerical integration shows that our expansion works well in the region above the highest physical threshold. (orig.)

  5. Determination of the two-loop Lamb shift in lithiumlike bismuth

    International Nuclear Information System (INIS)

    Sapirstein, J.; Cheng, K. T.

    2001-01-01

    The energy levels of lithiumlike bismuth are shown to be accurately described in a representation-independent manner when all diagrams involving one and two photons, with the exception of the two-loop Lamb shift, are evaluated. Comparison with the experimental value of the 2p 3/2 -2s 1/2 splitting then shows that, assuming three-photon effects are negligible, the contribution of the two-loop Lamb shift is 0.175(39) eV

  6. New class of two-loop neutrino mass models with distinguishable phenomenology

    Science.gov (United States)

    Cao, Qing-Hong; Chen, Shao-Long; Ma, Ernest; Yan, Bin; Zhang, Dong-Ming

    2018-04-01

    We discuss a new class of neutrino mass models generated in two loops, and explore specifically three new physics scenarios: (A) doubly charged scalar, (B) dark matter, and (C) leptoquark and diquark, which are verifiable at the 14 TeV LHC Run-II. We point out how the different Higgs insertions will distinguish our two-loop topology with others if the new particles in the loop are in the simplest representations of the SM gauge group.

  7. Higgs-Boson Two-Loop Contributions to Electric Dipole Moments in the MSSM

    CERN Document Server

    Pilaftsis, Apostolos

    1999-01-01

    The complete set of Higgs-boson two-loop contributions to electric dipole moments of the electron and neutron is calculated in the minimal supersymmetric standard model. The electric dipole moments are induced by CP-violating trilinear couplings of the `CP-odd' and charged Higgs bosons to the scalar top and bottom quarks. Numerical estimates of the individual two-loop contributions to electric dipole moments are given.

  8. Landau singularities and symbology: one- and two-loop MHV amplitudes in SYM theory

    Energy Technology Data Exchange (ETDEWEB)

    Dennen, Tristan; Spradlin, Marcus; Volovich, Anastasia [Department of Physics, Brown University,Providence RI 02912 (United States)

    2016-03-14

    We apply the Landau equations, whose solutions parameterize the locus of possible branch points, to the one- and two-loop Feynman integrals relevant to MHV amplitudes in planar N=4 super-Yang-Mills theory. We then identify which of the Landau singularities appear in the symbols of the amplitudes, and which do not. We observe that all of the symbol entries in the two-loop MHV amplitudes are already present as Landau singularities of one-loop pentagon integrals.

  9. Landau singularities and symbology: one- and two-loop MHV amplitudes in SYM theory

    International Nuclear Information System (INIS)

    Dennen, Tristan; Spradlin, Marcus; Volovich, Anastasia

    2016-01-01

    We apply the Landau equations, whose solutions parameterize the locus of possible branch points, to the one- and two-loop Feynman integrals relevant to MHV amplitudes in planar N=4 super-Yang-Mills theory. We then identify which of the Landau singularities appear in the symbols of the amplitudes, and which do not. We observe that all of the symbol entries in the two-loop MHV amplitudes are already present as Landau singularities of one-loop pentagon integrals.

  10. The Conceptual Design of Innovative Safe PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Han-Gon [Centural Research Institute, Daejeon (Korea, Republic of); Heo, Sun [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2016-10-15

    Most of countries operating NPPs have been performed post-Fukushima improvements as short-term countermeasure to enhance the safety of operating NPPs. Separately, vendors have made efforts on developing passive safety systems as long-term and ultimate countermeasures. AP1000 designed by Westinghouse Electric Company has passive safety systems including the passive emergency core cooling system (PECCS), the passive residual heat removal system (PRHRS), and the passive containment cooling system (PCCS). ESBWR designed by GE-Hitachi also has passive safety systems consisting of the isolation condenser system, the gravity driven cooling system and the PCCS. Other countries including China and Russia have made efforts on developing passive safety systems for enhancing the safety of their plants. In this paper, we summarize the design goals and main design feature of innovative safe PWR, iPOWER which is standing for Innovative Passive Optimized World-wide Economical Reactor, and show the developing status and results of research projects. To mitigate an accident without electric power and enhance the safety level of PWR, the conceptual designs of passive safety system and innovative safe PWR have been performed. It includes the PECCS for core cooling and the PCCS for containment cooling. Now we are performing the small scale and separate effect tests for the PECCS and the PCCS and preparing the integral effect test for the PECCS and real scale test for the PCCS.

  11. Application of the integrated analysis of safety (IAS) to sequences of Total loss of feed water in a PWR Reactor; Aplicacion del Analisis Integrado de Seguridad (ISA) a Secuencias de Perdidas Total de Agua de Alimentacion en un Reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Moreno Chamorro, P.; Gallego Diaz, C.

    2011-07-01

    The main objective of this work is to show the current status of the implementation of integrated analysis of safety (IAS) methodology and its SCAIS associated tool (system of simulation codes for IAS) to the sequence analysis of total loss of feedwater in a PWR reactor model Westinghouse of three loops with large, dry containment.

  12. Application of the integrated analysis of safety (ISA) to sequences of Total loss of feed water in a PWR Reactor

    International Nuclear Information System (INIS)

    Moreno Chamorro, P.; Gallego Diaz, C.

    2011-01-01

    The main objective of this work is to show the current status of the implementation of integrated analysis of safety (ISA) methodology and its SCAIS associated tool (system of simulation codes for ISA) to the sequence analysis of total loss of feedwater in a PWR reactor model Westinghouse of three loops with large, dry containment.

  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. 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)

  15. 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)

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

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

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

  17. Two-loop mass splittings in electroweak multiplets: Winos and minimal dark matter

    Science.gov (United States)

    McKay, James; Scott, Pat

    2018-03-01

    The radiatively-induced splitting of masses in electroweak multiplets is relevant for both collider phenomenology and dark matter. Precision two-loop corrections of O (MeV ) to the triplet mass splitting in the wino limit of the minimal supersymmetric standard model can affect particle lifetimes by up to 40%. We improve on previous two-loop self-energy calculations for the wino model by obtaining consistent input parameters to the calculation via two-loop renormalization-group running, and including the effect of finite light quark masses. We also present the first two-loop calculation of the mass splitting in an electroweak fermionic quintuplet, corresponding to the viable form of minimal dark matter (MDM). We place significant constraints on the lifetimes of the charged and doubly-charged fermions in this model. We find that the two-loop mass splittings in the MDM quintuplet are not constant in the large-mass limit, as might naively be expected from the triplet calculation. This is due to the influence of the additional heavy fermions in loop corrections to the gauge boson propagators.

  18. SQED two-loop beta function in the context of Implicit regularization

    International Nuclear Information System (INIS)

    Cherchiglia, Adriano Lana; Sampaio, Marcos; Nemes, Maria Carolina

    2013-01-01

    Full text: In this work we present the state-of-art for Implicit Regularization (IReg) in the context of supersymmetric theories. IReg is a four-dimensional regularization technique in momentum space which disentangles, in a consistent way at arbitrary order, the divergencies, regularization dependent and finite parts of any Feynman amplitude. Since it does not resort to modifications on the physical space-time dimensions of the underlying quantum field theoretical model, it can be consistently applied to supersymmetric theories. First we describe the technique and present previous results for supersymmetric models: the two-loop beta function for the Wess-Zumino model (both in the component and superfield formalism); the two-loop beta function for Super Yang-Mills (in the superfield formalism using the background field technique). After, we present our calculation of the two-loop beta function for massless and massive SQED using the superfield formalism with and without resorting to the background field technique. We find that only in the second case the two-loop divergence cancels out. We argue it is due to an anomalous Jacobian under the rescaling of the fields in the path-integral which is necessary for the application of the supersymmetric background field technique. We find, however, that in both cases the two-loop coefficients of beta function are non-null. Finally we briefly discuss the anomaly puzzle in the context of our technique. (author)

  19. Development of Cost Estimation Methodology of Decommissioning for PWR

    International Nuclear Information System (INIS)

    Lee, Sang Il; Yoo, Yeon Jae; Lim, Yong Kyu; Chang, Hyeon Sik; Song, Geun Ho

    2013-01-01

    The permanent closure of nuclear power plant should be conducted with the strict laws and the profound planning including the cost and schedule estimation because the plant is very contaminated with the radioactivity. In Korea, there are two types of the nuclear power plant. One is the pressurized light water reactor (PWR) and the other is the pressurized heavy water reactor (PHWR) called as CANDU reactor. Also, the 50% of the operating nuclear power plant in Korea is the PWRs which were originally designed by CE (Combustion Engineering). There have been experiences about the decommissioning of Westinghouse type PWR, but are few experiences on that of CE type PWR. Therefore, the purpose of this paper is to develop the cost estimation methodology and evaluate technical level of decommissioning for the application to CE type PWR based on the system engineering technology. The aim of present study is to develop the cost estimation methodology of decommissioning for application to PWR. Through the study, the following conclusions are obtained: · Based on the system engineering, the decommissioning work can be classified as Set, Subset, Task, Subtask and Work cost units. · The Set and Task structure are grouped as 29 Sets and 15 Task s, respectively. · The final result shows the cost and project schedule for the project control and risk management. · The present results are preliminary and should be refined and improved based on the modeling and cost data reflecting available technology and current costs like labor and waste data

  20. Two-Loop Correction to the Higgs Boson Mass in the MRSSM

    International Nuclear Information System (INIS)

    Stöckinger, Dominik; Diessner, Philip; Kotlarski, Wojciech; Kalinowski, Jan

    2015-01-01

    We present the impact of two-loop corrections on the mass of the lightest Higgs boson in the minimal R-symmetric supersymmetric standard model (MRSSM). These shift the Higgs boson mass up by typically 5 GeV or more. The dominant corrections arise from strong interactions, and from the gluon and its N=2 superpartners, the sgluon and Dirac gluino, and these corrections further increase with large Dirac gluino mass. The two-loop contributions governed purely by Yukawa couplings and the MRSSM λ, Λ parameters are smaller. We also update our earlier analysis which showed that the MRSSM can accommodate the measured Higgs and W boson masses. Including the two-loop corrections increases the parameter space where the theory prediction agrees with the measurement.

  1. Two-loop renormalization in the standard model, part I. Prolegomena

    Energy Technology Data Exchange (ETDEWEB)

    Actis, S. [Deutsches Elektronen-Synchrotron (DESY), Zeuthen (Germany); Ferroglia, A. [Albert-Ludwigs-Univ., Freiburg (Germany). Fakultat fur Phys.]|[Zuerich Univ. (Switzerland). Inst. fuer Theoretische Physik; Passera, M. [Padua Univ. (Italy). Dipt. di Fisica]|[INFN, Sezione di Padova (Italy); Passarino, G. [Torino Univ. (Italy). Dipt. di Fisica Teorica]|[INFN, Sezione di Torino (Italy)

    2006-12-15

    In this paper the building blocks for the two-loop renormalization of the Standard Model are introduced with a comprehensive discussion of the special vertices induced in the Lagrangian by a particular diagonalization of the neutral sector and by two alternative treatments of the Higgs tadpoles. Dyson resummed propagators for the gauge bosons are derived, and two-loop Ward-Slavnov-Taylor identities are discussed. In part II, the complete set of counterterms needed for the two-loop renormalization will be derived. In part III, a renormalization scheme will be introduced, connecting the renormalized quantities to an input parameter set of (pseudo-)experimental data, critically discussing renormalization of a gauge theory with unstable particles. (orig.)

  2. N ≥ 4 Supergravity Amplitudes from Gauge Theory at Two Loops

    International Nuclear Information System (INIS)

    Boucher-Veronneau, Camille

    2012-01-01

    We present the full two-loop four-graviton amplitudes in N = 4, 5, 6 supergravity. These results were obtained using the double-copy structure of gravity, which follows from the recently conjectured color-kinematics duality in gauge theory. The two-loop four-gluon scattering amplitudes in N = 0, 1, 2 supersymmetric gauge theory are a second essential ingredient. The gravity amplitudes have the expected infrared behavior: the two-loop divergences are given in terms of the squares of the corresponding one-loop amplitudes. The finite remainders are presented in a compact form. The finite remainder for N = 8 supergravity is also presented, in a form that utilizes a pure function with a very simple symbol.

  3. Two-loop operator matrix elements for massive fermionic local twist-2 operators in QED

    International Nuclear Information System (INIS)

    Bluemlein, J.; Freitas, A. de; Universidad Simon Bolivar, Caracas; Neerven, W.L. van

    2011-11-01

    We describe the calculation of the two--loop massive operator matrix elements with massive external fermions in QED. We investigate the factorization of the O(α 2 ) initial state corrections to e + e - annihilation into a virtual boson for large cms energies s >>m 2 e into massive operator matrix elements and the massless Wilson coefficients of the Drell-Yan process adapting the color coefficients to the case of QED, as proposed by F. A. Berends et. al. (Nucl. Phys. B 297 (1988)429). Our calculations show explicitly that the representation proposed there works at one-loop order and up to terms linear in ln (s/m 2 e ) at two-loop order. However, the two-loop constant part contains a few structural terms, which have not been obtained in previous direct calculations. (orig.)

  4. Two-loop corrections for nuclear matter in the Walecka model

    International Nuclear Information System (INIS)

    Furnstahl, R.J.; Perry, R.J.; Serot, B.D.; Department of Physics, The Ohio State University, Columbus, Ohio 43210; Physics Department and Nuclear Theory Center, Indiana University, Bloomington, Indiana 47405)

    1989-01-01

    Two-loop corrections for nuclear matter, including vacuum polarization, are calculated in the Walecka model to study the loop expansion as an approximation scheme for quantum hadrodynamics. Criteria for useful approximation schemes are discussed, and the concepts of strong and weak convergence are introduced. The two-loop corrections are evaluated first with one-loop parameters and mean fields and then by minimizing the total energy density with respect to the scalar field and refitting parameters to empirical nuclear matter saturation properties. The size and nature of the corrections indicate that the loop expansion is not convergent at two-loop order in either the strong or weak sense. Prospects for alternative approximation schemes are discussed

  5. The two-loop master integrals for qq-bar→VV

    International Nuclear Information System (INIS)

    Gehrmann, Thomas; Manteuffel, Andreas von; Tancredi, Lorenzo; Weihs, Erich

    2014-01-01

    We compute the full set of two-loop Feynman integrals appearing in massless two-loop four-point functions with two off-shell legs with the same invariant mass. These integrals allow to determine the two-loop corrections to the amplitudes for vector boson pair production at hadron colliders, qq-bar→VV, and thus to compute this process to next-to-next-to-leading order accuracy in QCD. The master integrals are derived using the method of differential equations, employing a canonical basis for the integrals. We obtain analytical results for all integrals, expressed in terms of multiple polylogarithms. We optimize our results for numerical evaluation by employing functions which are real valued for physical scattering kinematics and allow for an immediate power series expansion

  6. Color ferromagnetic vacuum states in QCD and two-loop energy densities

    International Nuclear Information System (INIS)

    Nielsen, H.B.; Ninomiya, M.

    1979-12-01

    Two-loop energy densities of color ferromagnetic states are obtained using the β-function calculated to two-loop approximation and the exact formula for the energy density of such a state. This is used to derive bounds on the MIT bag constant correcting the previous bound in one-loop approximation. For a constant field color ferromagnetic ansatz state the bound on the QCD scale parameter Λsub(p) 3 -vacuum ansatz with two-loop and instanton correction gives Λsub(p)<= 0.16 GeV. Tt is stressed that the 'perturbative vacuum', which is identified with the inside bag state is a somewhat ill defined concept due to a path-dependence in the integral giving the energy density. (Auth.)

  7. Analytic continuation of massless two-loop four-point functions

    International Nuclear Information System (INIS)

    Gehrmann, T.; Remiddi, E.

    2002-01-01

    We describe the analytic continuation of two-loop four-point functions with one off-shell external leg and internal massless propagators from the Euclidean region of space-like 1→3 decay to Minkowskian regions relevant to all 1→3 and 2→2 reactions with one space-like or time-like off-shell external leg. Our results can be used to derive two-loop master integrals and unrenormalized matrix elements for hadronic vector-boson-plus-jet production and deep inelastic two-plus-one-jet production, from results previously obtained for three-jet production in electron-positron annihilation. (author)

  8. Two-Loop Self-Energy Correction in a Strong Coulomb Nuclear Field

    International Nuclear Information System (INIS)

    Yerokhin, V.A.; Indelicato, P.; Shabaev, V.M.

    2005-01-01

    The two-loop self-energy correction to the ground-state energy levels of hydrogen-like ions with nuclear charges Z ≥ 10 is calculated without the Zα expansion, where α is the fine-structure constant. The data obtained are compared with the results of analytical calculations within the Zα expansion; significant disagreement with the analytical results of order α 2 (Zα) 6 has been found. Extrapolation is used to obtain the most accurate value for the two-loop self-energy correction for the 1s state in hydrogen

  9. Exact two-loop vacuum polarization correction to the Lamb shift in hydrogenlike ions

    International Nuclear Information System (INIS)

    Plunien, G.; Beier, T.; Soff, G.

    1998-01-01

    We present a calculation scheme for the two-loop vacuum polarization correction of order α 2 to the Lamb shift of hydrogenlike high-Z atoms. The interaction with the external Coulomb field is taken into account to all orders in (Zα). By means of a modified potential approach the problem is reduced to the evaluation of effective one-loop vacuumpolarization potentials. An expression for the energy shift is deduced within the framework of partial wave decomposition performing appropriate subtractions. Exact results for the two-loop vacuum polarization contribution to the Lamb shift of K- and L-shell electron states in hydrogenlike lead and uranium are presented. (orig.)

  10. Two loop effective Kahler potential of (non)-renormalizable supersymmetric models

    International Nuclear Information System (INIS)

    Groot Nibbelink, S.; Nyawelo, T.S.

    2005-10-01

    We perform a supergraph computation of the effective Kahler potential at one and two loops for general four dimensional N=1 supersymmetric theories described by arbitrary Kahler potential, superpotential and gauge kinetic function. We only insist on gauge invariance of the Kahler potential and the superpotential as we heavily rely on its consequences in the quantum theory. However, we do not require gauge invariance for the gauge kinetic functions, so that our results can also be applied to anomalous theories that involve the Green-Schwarz mechanism. We illustrate our two loop results by considering a few simple models: the (non-)renormalizable Wess-Zumino model and Super Quantum Electrodynamics. (author)

  11. Load-following operation of PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Jong Hwa; Oh, Soo Yul; Koo, Yang Hyun; Lee, Jae Han [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1993-12-01

    The load-following operation of nuclear power plants will become inevitable due to the increased nuclear share in the total electricity generation. As a groundwork for the load-following capability of the Korean next generation PWRs, the state-of-the-art has been reviewed. The core control principles and methods are the main subject in this review as well as the impact of load-following operations on the fuel performance and on the mechanical integrity of components. To begin with, it was described what the load-following operation is and in what view point the technology should be reviewed. Afterwards the load-following method, performance and problems in domestic 900 MWe class PWRs were discussed, and domestic R and D works were summarized. Foreign technologies were also reviewed. They include Mode G and Mode X of Foratom, D and L bank method of KWU, the method using PSCEA of ABB-CE, and MSHIM of Westinghouse. The load-following related special features of Foratom`s N4 plant, KWU`s plants, ABB-CE`s Systems 80+, and Westinghouse`s AP600 were described in each technology review. The review concluded that the capability of N4 plant with Mode X is the best and the methods in System, 80+ and AP600 would require verifications for the continued and usual load-following operation. It was recommended that the load-following operation experiences in domestic PWRs under operation be required to settle down the capability for the future. In addition, a more enhanced technology is required for the Korean next generation PWR regardless what the reference plant concept is. 30 figs., 19 tabs., 75 refs. (Author).

  12. 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

  13. Maturity of the PWR

    International Nuclear Information System (INIS)

    Bacher, P.; Rapin, M.; Aboudarham, L.; Bitsch, D.

    1983-03-01

    Figures illustrating the predominant position of the PWR system are presented. The question is whether on the basis of these figures the PWR can be considered to have reached maturity. The following analysis, based on the French program experience, is an attempt to pinpoint those areas in which industrial maturity of the PWR has been attained, and in which areas a certain evolution can still be expected to take place

  14. System-Level Heat Transfer Analysis, Thermal- Mechanical Cyclic Stress Analysis, and Environmental Fatigue Modeling of a Two-Loop Pressurized Water Reactor. A Preliminary Study

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish [Argonne National Lab. (ANL), Argonne, IL (United States); Soppet, William [Argonne National Lab. (ANL), Argonne, IL (United States); Majumdar, Saurin [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, Ken [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-01-03

    This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in April 2015 under the work package for environmentally assisted fatigue under DOE's Light Water Reactor Sustainability program. In this report, updates are discussed related to a system level preliminary finite element model of a two-loop pressurized water reactor (PWR). Based on this model, system-level heat transfer analysis and subsequent thermal-mechanical stress analysis were performed for typical design-basis thermal-mechanical fatigue cycles. The in-air fatigue lives of components, such as the hot and cold legs, were estimated on the basis of stress analysis results, ASME in-air fatigue life estimation criteria, and fatigue design curves. Furthermore, environmental correction factors and associated PWR environment fatigue lives for the hot and cold legs were estimated by using estimated stress and strain histories and the approach described in NUREG-6909. The discussed models and results are very preliminary. Further advancement of the discussed model is required for more accurate life prediction of reactor components. This report only presents the work related to finite element modelling activities. However, in between multiple tensile and fatigue tests were conducted. The related experimental results will be presented in the year-end report.

  15. 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

  16. A study on thimble plug removal for PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Song, Dong Soo; Lee, Chang Sup; Lee, Jae Yong; Jun, Hwang Yong [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    The thermal-hydraulic effects of removing the RCC guide thimble plugs are evaluated for 8 Westinghouse type PWR plants in Korea as a part of feasibility study: core outlet loss coefficient, thimble bypass flow, and best estimate flow. It is resulted that the best estimate thimble bypass flow increases about by 2% and the best estimate flow increases approximately by 1.2%. The resulting DNBR penalties can be covered with the current DNBR margin. Accident analyses are also investigated that the dropped rod transient is shown to be limiting and relatively sensitive to bypass flow variation. 8 refs., 5 tabs. (Author)

  17. A study on thimble plug removal for PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Song, Dong Soo; Lee, Chang Sup; Lee, Jae Yong; Jun, Hwang Yong [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    The thermal-hydraulic effects of removing the RCC guide thimble plugs are evaluated for 8 Westinghouse type PWR plants in Korea as a part of feasibility study: core outlet loss coefficient, thimble bypass flow, and best estimate flow. It is resulted that the best estimate thimble bypass flow increases about by 2% and the best estimate flow increases approximately by 1.2%. The resulting DNBR penalties can be covered with the current DNBR margin. Accident analyses are also investigated that the dropped rod transient is shown to be limiting and relatively sensitive to bypass flow variation. 8 refs., 5 tabs. (Author)

  18. 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%

  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. 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

  1. Two-loop anomalous dimensions for four-Fermi operators in supersymmetric theories

    Directory of Open Access Journals (Sweden)

    Junji Hisano

    2017-09-01

    Full Text Available We derive two-loop anomalous dimensions for four-Fermi operators in supersymmetric theories using the effective Kähler potential. We introduce the general forms in generic gauge theories and apply our results to the flavor-changing operators in (minimal supersymmetric standard models.

  2. The width of the Δ-resonance at two loop order in baryon chiral perturbation theory

    Energy Technology Data Exchange (ETDEWEB)

    Gegelia, Jambul, E-mail: j.gegelia@fz-juelich.de [Institute for Advanced Simulation, Institut für Kernphysik and Jülich Center for Hadron Physics, Forschungszentrum Jülich, D-52425 Jülich (Germany); Tbilisi State University, 0186 Tbilisi, Georgia (United States); Meißner, Ulf-G., E-mail: meissner@hiskp.uni-bonn.de [Helmholtz Institut für Strahlen- und Kernphysik and Bethe Center for Theoretical Physics, Universität Bonn, D-53115 Bonn (Germany); Institute for Advanced Simulation, Institut für Kernphysik and Jülich Center for Hadron Physics, Forschungszentrum Jülich, D-52425 Jülich (Germany); Siemens, Dmitrij, E-mail: dmitrij.siemens@rub.de [Institut für Theoretische Physik II, Ruhr-Universität Bochum, D-44780 Bochum (Germany); Yao, De-Liang, E-mail: d.yao@fz-juelich.de [Institute for Advanced Simulation, Institut für Kernphysik and Jülich Center for Hadron Physics, Forschungszentrum Jülich, D-52425 Jülich (Germany)

    2016-12-10

    We calculate the width of the delta resonance at leading two-loop order in baryon chiral perturbation theory. This gives a correlation between the leading pion–nucleon–delta and pion–delta couplings, which is relevant for the analysis of pion–nucleon scattering and other processes.

  3. The SU(2|3) dynamic two-loop form factors

    International Nuclear Information System (INIS)

    Brandhuber, A.; Kostacińska, M.; Penante, B.; Travaglini, G.; Young, D.

    2016-01-01

    We compute two-loop form factors of operators in the SU(2|3) closed subsector of N = 4 supersymmetric Yang-Mills. In particular, we focus on the non-protected, dimension-three operators Tr(X[Y,Z]) and Tr(ψψ) for which we compute the four possible two-loop form factors, and corresponding remainder functions, with external states 〈X̄ȲZ̄| and 〈ψ̄ψ̄|. Interestingly, the maximally transcendental part of the two-loop remainder of 〈X̄ȲZ̄|Tr(X[Y,Z])|0〉 turns out to be identical to that of the corresponding known quantity for the half-BPS operator Tr(X"3). We also find a surprising connection between the terms subleading in transcendentality and certain a priori unrelated remainder densities introduced in the study of the spin chain Hamiltonian in the SU(2) sector. Next, we use our calculation to resolve the mixing, recovering anomalous dimensions and eigenstates of the dilatation operator in the SU(2|3) sector at two loops. We also speculate on potential connections between our calculations in N = 4 super Yang-Mills and Higgs + multi-gluon amplitudes in QCD in an effective Lagrangian approach.

  4. The SU(2|3) dynamic two-loop form factors

    Energy Technology Data Exchange (ETDEWEB)

    Brandhuber, A.; Kostacińska, M. [Centre for Research in String Theory, School of Physics and Astronomy,Queen Mary University of London,Mile End Road, London E1 4NS (United Kingdom); Penante, B. [Centre for Research in String Theory, School of Physics and Astronomy,Queen Mary University of London,Mile End Road, London E1 4NS (United Kingdom); Institut für Physik und IRIS Adlershof, Humboldt Universität zu Berlin,Zum Großen Windkanal 6, 12489 Berlin (Germany); Travaglini, G.; Young, D. [Centre for Research in String Theory, School of Physics and Astronomy,Queen Mary University of London,Mile End Road, London E1 4NS (United Kingdom)

    2016-08-23

    We compute two-loop form factors of operators in the SU(2|3) closed subsector of N = 4 supersymmetric Yang-Mills. In particular, we focus on the non-protected, dimension-three operators Tr(X[Y,Z]) and Tr(ψψ) for which we compute the four possible two-loop form factors, and corresponding remainder functions, with external states 〈X̄ȲZ̄| and 〈ψ̄ψ̄|. Interestingly, the maximally transcendental part of the two-loop remainder of 〈X̄ȲZ̄|Tr(X[Y,Z])|0〉 turns out to be identical to that of the corresponding known quantity for the half-BPS operator Tr(X{sup 3}). We also find a surprising connection between the terms subleading in transcendentality and certain a priori unrelated remainder densities introduced in the study of the spin chain Hamiltonian in the SU(2) sector. Next, we use our calculation to resolve the mixing, recovering anomalous dimensions and eigenstates of the dilatation operator in the SU(2|3) sector at two loops. We also speculate on potential connections between our calculations in N = 4 super Yang-Mills and Higgs + multi-gluon amplitudes in QCD in an effective Lagrangian approach.

  5. Two-loop calculation of the effective potential for the Wess-Zumino model

    International Nuclear Information System (INIS)

    Fogleman, G.; Starkmann, G.D.; Viswanathan, K.S.; Simon Fraser Univ., Burnaby, British Columbia

    1983-01-01

    The effective potential for the supersymmetric Wess-Zumino model is computed off-shell to two loops. A renormalization procedure which preserves positivity of the kinetic terms in the effective action is implemented. Supersymmetry is not broken to this order. (orig.)

  6. The complete two-loop integrated jet thrust distribution in soft-collinear effective theory

    International Nuclear Information System (INIS)

    Manteuffel, Andreas von; Schabinger, Robert M.; Zhu, Hua Xing

    2014-01-01

    In this work, we complete the calculation of the soft part of the two-loop integrated jet thrust distribution in e + e − annihilation. This jet mass observable is based on the thrust cone jet algorithm, which involves a veto scale for out-of-jet radiation. The previously uncomputed part of our result depends in a complicated way on the jet cone size, r, and at intermediate stages of the calculation we actually encounter a new class of multiple polylogarithms. We employ an extension of the coproduct calculus to systematically exploit functional relations and represent our results concisely. In contrast to the individual contributions, the sum of all global terms can be expressed in terms of classical polylogarithms. Our explicit two-loop calculation enables us to clarify the small r picture discussed in earlier work. In particular, we show that the resummation of the logarithms of r that appear in the previously uncomputed part of the two-loop integrated jet thrust distribution is inextricably linked to the resummation of the non-global logarithms. Furthermore, we find that the logarithms of r which cannot be absorbed into the non-global logarithms in the way advocated in earlier work have coefficients fixed by the two-loop cusp anomalous dimension. We also show that in many cases one can straightforwardly predict potentially large logarithmic contributions to the integrated jet thrust distribution at L loops by making use of analogous contributions to the simpler integrated hemisphere soft function

  7. The light-cone gauge at two loops: The scalar anomalous dimension

    International Nuclear Information System (INIS)

    Capper, D.M.; Suzuki, A.T.; Jones, D.R.T.

    1985-01-01

    We demonstrate that the light-cone gauge is a feasible tool for multi-loop computations by using it to evaluate the two-loop scalar anomalous dimension, γsup((2)), in a general gauge theory. In the special case of supersymmetry we obtain agreement with previous results which were derived using nonlight-cone techniques. (orig.)

  8. A two-loop four-gluon helicity amplitude in QCD

    Energy Technology Data Exchange (ETDEWEB)

    Dixon, L.

    2000-01-06

    The authors present the two-loop pure gauge contribution to the gluon-gluon scattering amplitude with maximal helicity violation. The construction of the amplitude does not rely directly on Feynman diagrams, but instead uses its analytic properties 4--2{epsilon} dimensions. The authors evaluate the loop integrals appearing in the amplitude through order({epsilon}{sup 0})in terms of polylogarithms.

  9. Computation of Groebner bases for two-loop propagator type integrals

    International Nuclear Information System (INIS)

    Tarasov, O.V.

    2004-01-01

    The Groebner basis technique for calculating Feynman diagrams proposed in (Acta Phys. Pol. B 29(1998) 2655) is applied to the two-loop propagator type integrals with arbitrary masses and momentum. We describe the derivation of Groebner bases for all integrals with 1PI topologies and present explicit content of the Groebner bases

  10. Computation of Groebner bases for two-loop propagator type integrals

    Energy Technology Data Exchange (ETDEWEB)

    Tarasov, O.V. [DESY Zeuthen, Theory Group, Deutsches Elektronen Synchrotron, DESY, Platanenallee 6, D-15738 Zeuthen (Germany)]. E-mail: tarasov@ifh.de

    2004-11-21

    The Groebner basis technique for calculating Feynman diagrams proposed in (Acta Phys. Pol. B 29(1998) 2655) is applied to the two-loop propagator type integrals with arbitrary masses and momentum. We describe the derivation of Groebner bases for all integrals with 1PI topologies and present explicit content of the Groebner bases.

  11. Application of 't Hooft's renormalization scheme to two-loop calculations 230

    International Nuclear Information System (INIS)

    Vladimirov, A.A.

    1975-01-01

    The advantages of the Hooft scheme for asymptotic calculations in the renormalization group have been demonstrated. Two-loop calculations have been carried out in three renormalized models: in scalar electrodynamics, in a pseudoscalar Yukawa theory and in the Weiss-Zumino supersymmetrical model [ru

  12. Experiments on natural circulation during PWR severe accidents and their analysis

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Stewart, W.A.; Sha, W.T.

    1988-01-01

    Buoyancy-induced natural circulation flows will occur during the early-part of PWR high pressure accident scenarios. These flows affect several key parameters; in particular, the course of such accidents will most probably change due to local failures occurring in the primary coolant system (CS) before substantial core degradation. Natural circulation flow patterns were measured in a one-seventh scale PWR PCS facility at Westinghouse RandD laboratories. The measured flow and temperature distributions are report in this paper. The experiments were analyzed with the COMMIX code and good agreement was obtained between data and calculations. 10 refs., 8 figs., 2 tabs

  13. Electrical and control aspects of the Sizewell B PWR

    International Nuclear Information System (INIS)

    1992-01-01

    The pressurized water reactor, Sizewell-B, which is being built in Suffolk is well on in its construction schedule. This conference looked at the electrical and control aspects of the first PWR to be built in the United Kingdom. Although based on the standard Westinghouse PWR design, modifications have been made to meet the particular requirements of the site and the UK licensing regulations. There are 11 papers on all aspects of the electrical systems, 5 papers on the cables and cable installation, 5 on the main control rooms and auxiliary shutdown room, 5 on the integrated system and centralised operation, 6 on the monitoring and protection systems and 9 on the reactor protection systems. All 41 are indexed separately. (UK)

  14. 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.)

  15. 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

  16. Generation of SCALE 6 Input Data File for Cross Section Library of PWR Spent Fuel

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Cho, Dong Keun

    2010-11-01

    In order to obtain the cross section libraries of the Korean Pressurized water reactor (PWR) spent fuel (SF), SCALE 6 code input files have been generated. The PWR fuel data were obtained from the nuclear design report (NDR) of the current operating PWRs. The input file were prepared for 16 fuel types such as 4 types of Westinghouse 14x14, 3 types of OPR-1000 16x16, 4 types of Westinghouse 16x16, and 6 types of Westinghouse 17x17. For each fuel type, 5 kinds of fuel enrichments have been considered such as 1.5, 2.0 ,3.0, 4.0 and 5.0 wt%. In the SCALE 6 calculation, a ENDF-V 44 group was used. The 25 burnup step until 72000 MWD/T was used. A 1/4 symmetry model was used for 16x16 and 17x17 fuel assembly, and 1/2 symmetry model was used for 14x14 fuel assembly The generated cross section libraries will be used for the source-term analysis of the PWR SF

  17. Proposal for a advanced PWR core with adequate characteristics for passive safety concept

    International Nuclear Information System (INIS)

    Perrotta, Jose Augusto

    1999-01-01

    This work presents a discussion upon the suitable from an advanced PWR core, classified by the EPRI as 'Passive PWR' (advanced reactor with passive safety concept to power plants with less than 600 MW electrical power). The discussion upon the type of core is based on nuclear fuel engineering concepts. Discussion is made on type of fuel materials, structural materials, geometric shapes and manufacturing process that are suitable to produce fuel assemblies which give good performance for this type of reactors. The analysis is guided by the EPRI requirements for Advanced Light Water Reactor (ALWR). By means of comparison, the analysis were done to Angra 1 (old type of 600 MWe PWR class), and the design of the Westinghouse Advanced PWR-AP600. It was verified as a conclusion of this work that the modern PWR fuels are suitable for advanced PWR's Nevertheless, this work presents a technical alternative to this kind of fuel, still using UO 2 as fuel, but changing its cylindrical form of pellets and pin type fuel element to plane shape pallets and plate type fuel element. This is not a novelty fuel, since it was used in the 50's at Shippingport Reactor and as an advanced version by CEA of France in the 70's. In this work it is proposed a new mechanical assembly design for this fuel, which can give adequate safety and operational performance to the core of a 'Passive PWR'. (author)

  18. The muon magnetic moment in the 2HDM: complete two-loop result

    International Nuclear Information System (INIS)

    Cherchiglia, Adriano; Kneschke, Patrick; Stöckinger, Dominik; Stöckinger-Kim, Hyejung

    2017-01-01

    We study the 2HDM contribution to the muon anomalous magnetic moment a μ and present the complete two-loop result, particularly for the bosonic contribution. We focus on the Aligned 2HDM, which has general Yukawa couplings and contains the type I, II, X, Y models as special cases. The result is expressed with physical parameters: three Higgs boson masses, Yukawa couplings, two mixing angles, and one quartic potential parameter. We show that the result can be split into several parts, each of which has a simple parameter dependence, and we document their general behavior. Taking into account constraints on parameters, we find that the full 2HDM contribution to a μ can accommodate the current experimental value, and the complete two-loop bosonic contribution can amount to (2⋯4)×10 −10 , more than the future experimental uncertainty.

  19. Local integrand representations of all two-loop amplitudes in planar SYM

    International Nuclear Information System (INIS)

    Bourjaily, Jacob L.; Trnka, Jaroslav

    2015-01-01

    We use generalized unitarity at the integrand-level to directly construct local, manifestly dual-conformally invariant formulae for all two-loop scattering amplitudes in planar, maximally supersymmetric Yang-Mills theory (SYM). This representation separates contributions into manifestly finite and manifestly divergent terms — in a way that renders all infrared-safe observables (including ratio functions) calculable without any need for regulation. These results perfectly match the all-loop BCFW recursion relations, to which we provide a closed-form solution valid through two-loop-order. Finally, we describe and document a MATHEMATICA package which implements these results, available as part of this work’s source files on the arXiv.

  20. Resummed two-loop calculation of the disjoining pressure of a symmetric electrolyte soap film

    International Nuclear Information System (INIS)

    Dean, D.S.; Horgan, R.R.

    2004-01-01

    In this paper we consider the calculation of the disjoining pressure of a symmetric electrolytic soap film correct to two loops in perturbation theory. We show that the disjoining pressure is finite when the loop expansion is resummed using a cumulant expansion and requires no short distance cutoff in order to give a finite result. The loop expansion is resummed in terms of an expansion in g=l B /l D where l D is the Debye length and l B is the Bjerrum length. We show that there there is a nonanalytic contribution of order g ln(g). We also show that the two-loop correction is greater than the one-loop term at large film thicknesses suggesting a nonperturbative correction to the one-loop result in this limit

  1. Two-loop renormalization in the standard model, part III. Renormalization equations and their solutions

    International Nuclear Information System (INIS)

    Actis, S.; Passarino, G.

    2006-12-01

    In part I and II of this series of papers all elements have been introduced to extend, to two loops, the set of renormalization procedures which are needed in describing the properties of a spontaneously broken gauge theory. In this paper, the final step is undertaken and finite renormalization is discussed. Two-loop renormalization equations are introduced and their solutions discussed within the context of the minimal standard model of fundamental interactions. These equations relate renormalized Lagrangian parameters (couplings and masses) to some input parameter set containing physical (pseudo-)observables. Complex poles for unstable gauge and Higgs bosons are used and a consistent setup is constructed for extending the predictivity of the theory from the Lep1 Z-boson scale (or the Lep2 WW scale) to regions of interest for LHC and ILC physics. (orig.)

  2. Two-loop renormalization in the standard model, part III. Renormalization equations and their solutions

    Energy Technology Data Exchange (ETDEWEB)

    Actis, S. [Deutsches Elektronen-Synchrotron (DESY), Zeuthen (Germany); Passarino, G. [Torino Univ. (Italy). Dipt. di Fisica Teorica; INFN, Sezione di Torino (Italy)

    2006-12-15

    In part I and II of this series of papers all elements have been introduced to extend, to two loops, the set of renormalization procedures which are needed in describing the properties of a spontaneously broken gauge theory. In this paper, the final step is undertaken and finite renormalization is discussed. Two-loop renormalization equations are introduced and their solutions discussed within the context of the minimal standard model of fundamental interactions. These equations relate renormalized Lagrangian parameters (couplings and masses) to some input parameter set containing physical (pseudo-)observables. Complex poles for unstable gauge and Higgs bosons are used and a consistent setup is constructed for extending the predictivity of the theory from the Lep1 Z-boson scale (or the Lep2 WW scale) to regions of interest for LHC and ILC physics. (orig.)

  3. Matching the $D^{6}R^{4}$ interaction at two-loops

    CERN Document Server

    D'Hoker, Eric; Pioline, Boris; Russo, Rodolfo

    2015-01-01

    The coefficient of the $D^6 {\\cal R}^4$ interaction in the low energy expansion of the two-loop four-graviton amplitude in type II superstring theory is known to be proportional to the integral of the Zhang-Kawazumi (ZK) invariant over the moduli space of genus-two Riemann surfaces. We demonstrate that the ZK invariant is an eigenfunction with eigenvalue 5 of the Laplace-Beltrami operator in the interior of moduli space. Exploiting this result, we evaluate the integral of the ZK invariant explicitly, finding agreement with the value of the two-loop $D^6 {\\cal R}^4$ interaction predicted on the basis of S-duality and supersymmetry. A review of the current understanding of the $D^{2p} {\\cal R}^4$ interactions in type II superstring theory compactified on a torus $T^d$ with $p \\leq 3$ and $d \\leq 4$ is included.

  4. Low-energy effective action in two-dimensional SQED: a two-loop analysis

    Science.gov (United States)

    Samsonov, I. B.

    2017-07-01

    We study two-loop quantum corrections to the low-energy effective actions in N=(2,2) and N=(4,4) SQED on the Coulomb branch. In the latter model, the low-energy effective action is described by a generalized Kähler potential which depends on both chiral and twisted chiral superfields. We demonstrate that this generalized Kähler potential is one-loop exact and corresponds to the N=(4,4) sigma-model with torsion presented by Roček, Schoutens and Sevrin [1]. In the N=(2,2) SQED, the effective Kähler potential is not protected against higher-loop quantum corrections. The two-loop quantum corrections to this potential and the corresponding sigma-model metric are explicitly found.

  5. Automatic calculation of massive two-loop self-energies with XLOOPS

    International Nuclear Information System (INIS)

    Franzkowski, J.

    1997-01-01

    Within the program package XLOOPS it is possible to calculate self-energies up to the two-loop level for arbitrary massive particles. The program package -written in MAPLE (Char et al., Maple V Language Reference Manual (Springer, 1991); Char et al., Maple V Library Reference Manual (Springer, 1991)) - is designed to deal with the full tensor structure of the occurring integrals. This means that applications are not restricted to those cases where the reduction to scalars via equivalence theorem is allowed. The algorithms handle two-loop integrals analytically if this is possible. For those topologies where no analytic result for the general mass case is available, the diagrams are reduced to integral representations which encounter at most at two-fold integration. These integral representations are numerically stable and can be performed easily using VEGAS (Lepage, J. Comp. Phys. 27 (1978) 192; Lepage, Cornell Univ. Preprint CLNS-80/447 (1980)). (orig.)

  6. BPS Wilson loops and Bremsstrahlung function in ABJ(M): a two loop analysis

    Energy Technology Data Exchange (ETDEWEB)

    Bianchi, Marco S. [Institut für Physik, Humboldt-Universität zu Berlin,Newtonstraße 15, 12489 Berlin (Germany); Griguolo, Luca [Dipartimento di Fisica e Scienze della Terra, Università di Parmaand INFN Gruppo Collegato di Parma,Viale G.P. Usberti 7/A, 43100 Parma (Italy); Leoni, Matias [Physics Department, FCEyN-UBA & IFIBA-CONICETCiudad Universitaria, Pabellón I, 1428, Buenos Aires (Argentina); Penati, Silvia [Dipartimento di Fisica, Università di Milano-Bicoccaand INFN, Sezione di Milano-Bicocca,Piazza della Scienza 3, I-20126 Milano (Italy); Seminara, Domenico [Dipartimento di Fisica, Università di Firenzeand INFN Sezione di Firenze,via G. Sansone 1, 50019 Sesto Fiorentino (Italy)

    2014-06-19

    We study a family of circular BPS Wilson loops in N=6 super Chern-Simons-matter theories, generalizing the usual 1/2-BPS circle. The scalar and fermionic couplings depend on two deformation parameters and these operators can be considered as the ABJ(M) counterpart of the DGRT latitudes defined in N=4 SYM. We perform a complete two-loop analysis of their vacuum expectation value, discuss the appearance of framing-like phases and propose a general relation with cohomologically equivalent bosonic operators. We make an all-loop proposal for computing the Bremsstrahlung function associated to the 1/2-BPS cusp in terms of these generalized Wilson loops. When applied to our two-loop result it reproduces the known expression. Finally, we comment on the generalization of this proposal to the bosonic 1/6-BPS case.

  7. Two-Loop Master Integrals for $\\gamma^{*} \\to 3$ Jets the Non-Planar Topologies

    CERN Document Server

    Gehrmann, T

    2001-01-01

    The calculation of the two-loop corrections to the three-jet production rate and to event shapes in electron--positron annihilation requires the computation of a number of two-loop four-point master integrals with one off-shell and three on-shell legs. Up to now, only those master integrals corresponding to planar topologies were known. In this paper, we compute the yet outstanding non-planar master integrals by solving differential equations in the external invariants which are fulfilled by these master integrals. We obtain the master integrals as expansions in $\\e=(4-d)/2$, where $d$ is the space-time dimension. The fully analytic results are expressed in terms of the two-dimensional harmonic polylogarithms already introduced in the evaluation of the planar topologies.

  8. Heavy-quark production in gluon fusion at two loops in QCD

    International Nuclear Information System (INIS)

    Czakon, M.

    2007-07-01

    We present the two-loop virtual QCD corrections to the production of heavy quarks in gluon fusion. The results are exact in the limit when all kinematical invariants are large compared to the mass of the heavy quark up to terms suppressed by powers of the heavy-quark mass. Our derivation uses a simple relation between massless and massive QCD scattering amplitudes as well as a direct calculation of the massive amplitude at two loops. The results presented here together with those obtained previously for quark-quark scattering form important parts of the next-to-next-to-leading order QCD corrections to heavy-quark production in hadron-hadron collisions. (orig.)

  9. Two-loop Higgs mass calculations beyond the MSSM with SARAH and SPheno

    Energy Technology Data Exchange (ETDEWEB)

    Nickel, Kilian [Physikalisches Institut, Universitaet Bonn (Germany); Staub, Florian [Theory Division, CERN, Geneva (Switzerland); Goodsell, Mark [LPTHE, UPMC Univ. Paris 06 (France)

    2015-07-01

    We present a recent extension to the Mathematica package SARAH which allows for Higgs mass calculations at the two-loop level in a wide range of supersymmetric models beyond the MSSM. These calculations are based on the effective potential approach. For the numerical evaluation Fortran code for SPheno is generated by SARAH. This allows to predict the Higgs mass in more complicated SUSY theories with a similar precision as most state-of-the-art spectrum generators do for the MSSM.

  10. Uncertainty of the two-loop RG upper bound on the Higgs mass

    International Nuclear Information System (INIS)

    Pirogov, Yu.F.; Zenin, O.V.

    2003-01-01

    A modified criterion of the SM perturbative consistency is proposed. It is based on the analytic properties of the two-loop SM running couplings. Under the criterion adopted, the Higgs mass up to 380 GeV might not give rise to strong coupling prior to the Planck scale. This means that the light Higgs boson is possibly preferred for reasons other than the SM perturbative consistency, i.e., for reasons beyond the SM

  11. Two-loop finiteness of self-energies in higher-derivative SQED3

    Directory of Open Access Journals (Sweden)

    E.A. Gallegos

    2015-09-01

    Full Text Available In the N=1 superfield formalism, two higher-derivative kinetic operators (Lee–Wick operators are implemented into the standard three dimensional supersymmetric quantum electrodynamics (SQED3 for improving its ultraviolet behavior. It is shown in particular that the ghosts associated with these Lee–Wick operators allow to remove all ultraviolet divergences in the scalar and gauge self-energies at two-loop level.

  12. Uncertainty of the two-loop RG upper bound on the Higgs mass

    International Nuclear Information System (INIS)

    Pirogov, Yu.F.; Zenin, O.V.

    2003-01-01

    A modified criterion of the standard model perturbative consistency is proposed. It is based on the analytic properties of the two-loop standard model running couplings. Under the criterion adopted, the Higgs mass up to 380 GeV might not give rise to the strong coupling prior to the Planck scale. This means that light Higgs boson is possibly preferred for reasons other than the standard model perturbative consistency, i.e., for reasons beyond the standard model [ru

  13. Two-loop QED corrections to the Altarelli-Parisi splitting functions

    Energy Technology Data Exchange (ETDEWEB)

    Florian, Daniel de [International Center for Advanced Studies (ICAS), UNSAM,Campus Miguelete, 25 de Mayo y Francia (1650) Buenos Aires (Argentina); Sborlini, Germán F.R.; Rodrigo, Germán [Instituto de Física Corpuscular, Universitat de València,Consejo Superior de Investigaciones Científicas,Parc Científic, E-46980 Paterna, Valencia (Spain)

    2016-10-11

    We compute the two-loop QED corrections to the Altarelli-Parisi (AP) splitting functions by using a deconstructive algorithmic Abelianization of the well-known NLO QCD corrections. We present explicit results for the full set of splitting kernels in a basis that includes the leptonic distribution functions that, starting from this order in the QED coupling, couple to the partonic densities. Finally, we perform a phenomenological analysis of the impact of these corrections in the splitting functions.

  14. Two-loop current–current operator contribution to the non-leptonic QCD penguin amplitude

    Directory of Open Access Journals (Sweden)

    G. Bell

    2015-11-01

    Full Text Available The computation of direct CP asymmetries in charmless B decays at next-to-next-to-leading order (NNLO in QCD is of interest to ascertain the short-distance contribution. Here we compute the two-loop penguin contractions of the current–current operators Q1,2 and provide a first estimate of NNLO CP asymmetries in penguin-dominated b→s transitions.

  15. Two-loop effective potential for Wess-Zumino model using superfields

    International Nuclear Information System (INIS)

    Santos, R.P. dos; Srivastava, P.P.

    1989-01-01

    For the case of several interacting chiral superfields the propagators for the unconstrained superfield potentials in the 'shifted' theory, where the supersymmetry is explicity broken, are derived in a compact form. They are used to compute the one-loop effective potential in the general case, while a superfield calculation of the renormalized effective potential to two loops for the Wess-Zumino models is performed. (authors) [pt

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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

  1. CHIRON: a package for ChPT numerical results at two loops

    Energy Technology Data Exchange (ETDEWEB)

    Bijnens, Johan [Lund University, Department of Astronomy and Theoretical Physics, Lund (Sweden)

    2015-01-01

    This document describes the package CHIRON which includes two libraries, chiron itself and jbnumlib.chiron is a set of routines useful for two-loop numerical results in chiral perturbation theory (ChPT). It includes programs for the needed one- and two-loop integrals as well as routines to deal with the ChPT parameters. The present version includes everything needed for the masses, decay constants and quark-antiquark vacuum-expectation-values. An added routine calculates consistent values for the masses and decay constants when the pion and kaon masses are varied. In addition a number of finite volume results are included: one-loop tadpole integrals, two-loop sunset integrals and the results for masses and decay constants. The numerical routine library jbnumlib contains the numerical routines used in chiron. Many are to a large extent simple C++ versions of routines in the CERNLIB numerical library. Notable exceptions are the dilogarithm and the Jacobi theta function implementations. This paper describes what is included in CHIRON v0.50. (orig.)

  2. Two-loop renormalization in the standard model, part II. Renormalization procedures and computational techniques

    Energy Technology Data Exchange (ETDEWEB)

    Actis, S. [Deutsches Elektronen-Synchrotron (DESY), Zeuthen (Germany); Passarino, G. [Torino Univ. (Italy). Dipt. di Fisica Teorica; INFN, Sezione di Torino (Italy)

    2006-12-15

    In part I general aspects of the renormalization of a spontaneously broken gauge theory have been introduced. Here, in part II, two-loop renormalization is introduced and discussed within the context of the minimal Standard Model. Therefore, this paper deals with the transition between bare parameters and fields to renormalized ones. The full list of one- and two-loop counterterms is shown and it is proven that, by a suitable extension of the formalism already introduced at the one-loop level, two-point functions suffice in renormalizing the model. The problem of overlapping ultraviolet divergencies is analyzed and it is shown that all counterterms are local and of polynomial nature. The original program of 't Hooft and Veltman is at work. Finite parts are written in a way that allows for a fast and reliable numerical integration with all collinear logarithms extracted analytically. Finite renormalization, the transition between renormalized parameters and physical (pseudo-)observables, are discussed in part III where numerical results, e.g. for the complex poles of the unstable gauge bosons, are shown. An attempt is made to define the running of the electromagnetic coupling constant at the two-loop level. (orig.)

  3. Two-Loop Gluon to Gluon-Gluon Splitting Amplitudes in QCD

    International Nuclear Information System (INIS)

    Bern, Z.

    2004-01-01

    Splitting amplitudes are universal functions governing the collinear behavior of scattering amplitudes for massless particles. We compute the two-loop g → gg splitting amplitudes in QCD, N = 1, and N = 4 super-Yang-Mills theories, which describe the limits of two-loop n-point amplitudes where two gluon momenta become parallel. They also represent an ingredient in a direct x-space computation of DGLAP evolution kernels at next-to-next-to-leading order. To obtain the splitting amplitudes, we use the unitarity sewing method. In contrast to the usual light-cone gauge treatment, our calculation does not rely on the principal-value or Mandelstam-Leibbrandt prescriptions, even though the loop integrals contain some of the denominators typically encountered in light-cone gauge. We reduce the integrals to a set of 13 master integrals using integration-by-parts and Lorentz invariance identities. The master integrals are computed with the aid of differential equations in the splitting momentum fraction z. The ε-poles of the splitting amplitudes are consistent with a formula due to Catani for the infrared singularities of two-loop scattering amplitudes. This consistency essentially provides an inductive proof of Catani's formula, as well as an ansatz for previously-unknown 1/ε pole terms having non-trivial color structure. Finite terms in the splitting amplitudes determine the collinear behavior of finite remainders in this formula

  4. 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)

  5. 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)

  6. 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

  7. Regularization independent analysis of the origin of two loop contributions to N=1 Super Yang-Mills beta function

    Energy Technology Data Exchange (ETDEWEB)

    Fargnoli, H.G.; Sampaio, Marcos; Nemes, M.C. [Federal University of Minas Gerais, ICEx, Physics Department, P.O. Box 702, Belo Horizonte, MG (Brazil); Hiller, B. [Coimbra University, Faculty of Science and Technology, Physics Department, Center of Computational Physics, Coimbra (Portugal); Baeta Scarpelli, A.P. [Setor Tecnico-Cientifico, Departamento de Policia Federal, Lapa, Sao Paulo (Brazil)

    2011-05-15

    We present both an ultraviolet and an infrared regularization independent analysis in a symmetry preserving framework for the N=1 Super Yang-Mills beta function to two loop order. We show explicitly that off-shell infrared divergences as well as the overall two loop ultraviolet divergence cancel out, whilst the beta function receives contributions of infrared modes. (orig.)

  8. Regularization independent analysis of the origin of two loop contributions to N=1 Super Yang-Mills beta function

    International Nuclear Information System (INIS)

    Fargnoli, H.G.; Sampaio, Marcos; Nemes, M.C.; Hiller, B.; Baeta Scarpelli, A.P.

    2011-01-01

    We present both an ultraviolet and an infrared regularization independent analysis in a symmetry preserving framework for the N=1 Super Yang-Mills beta function to two loop order. We show explicitly that off-shell infrared divergences as well as the overall two loop ultraviolet divergence cancel out, whilst the beta function receives contributions of infrared modes. (orig.)

  9. 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

  10. The PWR cores management

    International Nuclear Information System (INIS)

    Barral, J.C.; Rippert, D.; Johner, J.

    2000-01-01

    During the meeting of the 25 january 2000, organized by the SFEN, scientists and plant operators in the domain of the PWR debated on the PWR cores management. The five first papers propose general and economic information on the PWR and also the fast neutron reactors chains in the electric power market: statistics on the electric power industry, nuclear plant unit management, the ITER project and the future of the thermonuclear fusion, the treasurer's and chairman's reports. A second part offers more technical papers concerning the PWR cores management: performance and optimization, in service load planning, the cores management in the other countries, impacts on the research and development programs. (A.L.B.)

  11. 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

  12. Reformulating the TBA equations for the quark anti-quark potential and their two loop expansion

    International Nuclear Information System (INIS)

    Bajnok, Zoltán; Balog, János; Correa, Diego H.; Hegedűs, Árpád; Massolo, Fidel I. Schaposnik; Tóth, Gábor Zsolt

    2014-01-01

    The boundary thermodynamic Bethe Ansatz (BTBA) equations introduced in http://dx.doi.org/10.1007/JHEP08(2012)134http://dx.doi.org/10.1007/JHEP10(2013)135 to describe the cusp anomalous dimension contain imaginary chemical potentials and singular boundary fugacities, which make its systematic expansion problematic. We propose an alternative formulation based on real chemical potentials and additional source terms. We expand our equations to double wrapping order and find complete agreement with the direct two-loop gauge theory computation of the cusp anomalous dimension

  13. Local integrands for two-loop all-plus Yang-Mills amplitudes

    International Nuclear Information System (INIS)

    Badger, Simon; Mogull, Gustav; Peraro, Tiziano

    2016-01-01

    We express the planar five- and six-gluon two-loop Yang-Mills amplitudes with all positive helicities in compact analytic form using D-dimensional local integrands that are free of spurious singularities. The integrand is fixed from on-shell tree amplitudes in six dimensions using D-dimensional generalised unitarity cuts. The resulting expressions are shown to have manifest infrared behaviour at the integrand level. We also find simple representations of the rational terms obtained after integration in 4−2ϵ dimensions.

  14. Reggeon field theory at D = 2 in two-loop approximation

    International Nuclear Information System (INIS)

    Eremyan, Sh.S.; Nazaryan, A.E.

    1982-01-01

    A general method of constructing an explicit representation is developed for the pomeron propagator in the presence of additional parameters, such as the pomeron production threshold xi 0 , momentum transfer K vector or the intercept shift delta 0 . The method is shown to be applicable in both one-loop and two-loop approximations. The obtained general formulae allow to consider the pomeron propagator in both asymptotic region and the region of the perturbation theory applicability. Besides, they provide the smooth matching of both these regions. The observed values are calculated, and the results connected with asymptotically high energies are discussed

  15. The two-loop symbol of all multi-Regge regions

    International Nuclear Information System (INIS)

    Bargheer, Till; Schomerus, Volker; Papathanasiou, Georgios

    2015-12-01

    We study the symbol of the two-loop n-gluon MHV amplitude for all Mandelstam regions in multi-Regge kinematics in N= 4 super Yang-Mills theory. While the number of distinct Mandelstam regions grows exponentially with n, the increase of independent symbols turns out to be merely quadratic. We uncover how to construct the symbols for any number of external gluons from just two building blocks which are naturally associated with the six- and seven-gluon amplitude, respectively. The second building block is entirely new, and in addition to its symbol, we also construct a prototype function that correctly reproduces all terms of maximal functional transcendentality.

  16. Local integrands for two-loop all-plus Yang-Mills amplitudes

    Energy Technology Data Exchange (ETDEWEB)

    Badger, Simon; Mogull, Gustav; Peraro, Tiziano [Higgs Centre for Theoretical Physics, School of Physics and Astronomy,The University of Edinburgh, James Clerk Maxwell Building,Peter Guthrie Tait Road, Edinburgh EH9 3FD (United Kingdom)

    2016-08-09

    We express the planar five- and six-gluon two-loop Yang-Mills amplitudes with all positive helicities in compact analytic form using D-dimensional local integrands that are free of spurious singularities. The integrand is fixed from on-shell tree amplitudes in six dimensions using D-dimensional generalised unitarity cuts. The resulting expressions are shown to have manifest infrared behaviour at the integrand level. We also find simple representations of the rational terms obtained after integration in 4−2ϵ dimensions.

  17. The two-loop symbol of all multi-Regge regions

    International Nuclear Information System (INIS)

    Bargheer, Till; Papathanasiou, Georgios; Schomerus, Volker

    2016-01-01

    We study the symbol of the two-loop n-gluon MHV amplitude for all Mandelstam regions in multi-Regge kinematics in N=4 super Yang-Mills theory. While the number of distinct Mandelstam regions grows exponentially with n, the increase of independent symbols turns out to be merely quadratic. We uncover how to construct the symbols for any number of external gluons from just two building blocks which are naturally associated with the six- and seven-gluon amplitude, respectively. The second building block is entirely new, and in addition to its symbol, we also construct a prototype function that correctly reproduces all terms of maximal functional transcendentality.

  18. A complete two-loop, five-gluon helicity amplitude in Yang-Mills theory

    International Nuclear Information System (INIS)

    Badger, Simon; Mogull, Gustav; Ochirov, Alexander; O’Connell, Donal

    2015-01-01

    We compute the integrand of the full-colour, two-loop, five-gluon scattering amplitude in pure Yang-Mills theory with all helicities positive, using generalized unitarity cuts. Tree-level BCJ relations, satisfied by amplitudes appearing in the cuts, allow us to deduce all the necessary non-planar information for the full-colour amplitude from known planar data. We present our result in terms of irreducible numerators, with colour factors derived from the multi-peripheral colour decomposition. Finally, the leading soft divergences are checked to reproduce the expected infrared behaviour.

  19. New perturbative upper bound on MH from fermionic Higgs decays at two loops

    International Nuclear Information System (INIS)

    Durand, L.; Kniehl, B.A.; Riesselmann, K.

    1993-09-01

    We present the dominant two-loop O (G F 2 M H 4 ) electroweak corrections to the fermionic decay widths of a high-mass Higgs boson in the Standard Model. The corrections are negative and quite significant, and are larger in magnitude than the one-loop electroweak corrections for M H > or ∼400 GeV. This indicates the onset of a breakdown of perturbation theory in the Higgs sector of the Standard Model at this surprisingly low value of the Higgs boson mass. (orig.)

  20. Improved two-loop beam energy stabilizer for an FN tandem accelerator

    International Nuclear Information System (INIS)

    Trainor, T.A.

    1981-01-01

    A detailed analysis of the properties of various elements in a two-loop voltage regulator for a tandem accelerator enabled design of an optimum system which reduces effective accelerating voltage noise below 100 V. Essential features of the new system are high-quality slit preamplifiers, careful attention to removal of extraneous noise sources, and proper shaping of frequency responses to maximize stable gains and ensure compatibility of the two control loops. The resultant beam energy stabilizer system is easy to operate, has well defined indicators for proper adjustment of operating parameters, and recovers reliably from beam interruptions

  1. Electroweak two-loop corrections to the effective weak mixing angle

    International Nuclear Information System (INIS)

    Awramik, Malgorzata; Czakon, Michal; Freitas, Ayres

    2006-01-01

    Recently exact results for the complete electroweak two-loop contributions to the effective weak mixing angle were published. This paper illustrates the techniques used for this computation, in particular the methods for evaluating the loop diagrams and the proper definition of Z-pole observables at next-to-next-to-leading order. Numerical results are presented in terms of simple parametrization formulae and compared in detail with a previous result of an expansion up to next-to-leading order in the top-quark mass. Finally, an estimate of the remaining theoretical uncertainties from unknown higher-order corrections is given

  2. A complete two-loop, five-gluon helicity amplitude in Yang-Mills theory

    Energy Technology Data Exchange (ETDEWEB)

    Badger, Simon; Mogull, Gustav; Ochirov, Alexander [Higgs Centre for Theoretical Physics, School of Physics and Astronomy, University of Edinburgh, Edinburgh EH9 3JZ, Scotland (United Kingdom); O’Connell, Donal [Higgs Centre for Theoretical Physics, School of Physics and Astronomy, University of Edinburgh, Edinburgh EH9 3JZ, Scotland (United Kingdom); Kavli Institute for Theoretical Physics, University of California, Santa Barbara, CA 93106-4030 (United States)

    2015-10-09

    We compute the integrand of the full-colour, two-loop, five-gluon scattering amplitude in pure Yang-Mills theory with all helicities positive, using generalized unitarity cuts. Tree-level BCJ relations, satisfied by amplitudes appearing in the cuts, allow us to deduce all the necessary non-planar information for the full-colour amplitude from known planar data. We present our result in terms of irreducible numerators, with colour factors derived from the multi-peripheral colour decomposition. Finally, the leading soft divergences are checked to reproduce the expected infrared behaviour.

  3. New two-loop contribution to electric dipole moment in supersymmetric theories

    CERN Document Server

    Chang, Darwin; Pilaftsis, Apostolos; Chang, Darwin; Keung, Wai-Yee; Pilaftsis, Apostolos

    1999-01-01

    We calculate a new type of two-loop contributions to the electric dipole moments of the electron and neutron in supersymmetric theories. The new contributions are originated from the potential CP violation in the trilinear couplings of the Higgs bosons to the scalar-top or the scalar-bottom quarks. These couplings were previously very weakly constrained. The electric dipole moments are induced through a mechanism analogous to that due to Barr and Zee. We find observable effects for a sizeable portion of the parameter space related to the third generation scalar-quarks in the minimal supersymmetric standard model which cannot be excluded by earlier considerations.

  4. Reduction of the N-component scalar model at the two-loop level

    International Nuclear Information System (INIS)

    Jakovac, A.

    1996-01-01

    Dimensional reduction of high temperature field theories improves IR features of their perturbative treatment. A crucial question is the following: What three-dimensional theory is representing the full system the most faithful way? A careful investigation of the induced three-dimensional counterterm structure of the finite temperature 4D O(N) symmetric scalar theory at the two-loop level leads to proposing the presence of nonlocal operators in the effective theory. A three-dimensional matching process is applied for the construction of the optimal local, superrenormalizable approximation. copyright 1996 The American Physical Society

  5. 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

  6. 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.

  7. 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

  8. Description of the two-loop RELAP5 model of the L-Reactor at the Savannah River Site

    International Nuclear Information System (INIS)

    Cozzuol, J.M.; Davis, C.B.

    1989-12-01

    A two-loop RELAP5 input model of the L-Reactor at the Savannah River Site (SRS) was developed to support thermal-hydraulic analysis of SRS reactors. The model was developed to economically evaluate potential design changes. The primary simplifications in the model were in the number of loops and the detail in the moderator tank. The six loops in the reactor were modeled with two loops, one representing a single loop and the other representing five combined loops. The model has undergone a quality assurance review. This report describes the two-loop model, its limitations, and quality assurance. 29 refs., 18 figs., 10 tabs

  9. Two-loop renormalization group analysis of supersymmetric SO(10) models with an intermediate scale

    International Nuclear Information System (INIS)

    Bastero-Gil, M.; Brahmachari, B.

    1996-03-01

    Two-loop evolutions of the gauge couplings in a class of intermediate scale supersymmetric SO(10) models including the effect of third generation Yukawa couplings are studied. The unification scale, the intermediate scale and the value of the unification gauge coupling in these models are calculated and the gauge boson mediated proton decay rates are estimated. In some cases the predicted proton lifetime turns out to be in the border-line of experimental limit. The predictions of the top quark mass, the mass ratio m b (m b )/m τ (m τ ) from the two-loop evolution of Yukawa couplings and the mass of the left handed neutrino via see-saw mechanism are summarized. The lower bounds on the ratio of the VEVs of the two low energy doublets (tan β) from the requirement of the perturbative unitarity of the top quark Yukawa coupling up to the grand unification scale are also presented. All the predictions have been compared with those of the one-step unified theory. (author). 33 refs, 5 figs, 1 tab

  10. The impact of two-loop effects on the scenario of MSSM Higgs alignment without decoupling

    Energy Technology Data Exchange (ETDEWEB)

    Haber, Howard E.; Stefaniak, Tim [University of California, Santa Cruz Institute for Particle Physics (SCIPP) and Department of Physics, Santa Cruz, CA (United States); Heinemeyer, Sven [Campus of International Excellence UAM+CSIC, Madrid (Spain); Universidad Autonoma de Madrid, Instituto de Fisica Teorica, (UAM/CSIC), Madrid (Spain); Instituto de Fisica de Cantabria (CSIC-UC), Santander (Spain)

    2017-11-15

    In multi-Higgs models, the properties of one neutral scalar state approximate those of the Standard Model (SM) Higgs boson in a limit where the corresponding scalar field is roughly aligned in field space with the scalar doublet vacuum expectation value. In a scenario of alignment without decoupling, a SM-like Higgs boson can be accompanied by additional scalar states whose masses are of a similar order of magnitude. In the Minimal Supersymmetric Standard Model (MSSM), alignment without decoupling can be achieved due to an accidental cancellation of tree-level and radiative loop-level effects. In this paper we assess the impact of the leading two-loop O(α{sub s}h{sub t}{sup 2}) corrections on the Higgs alignment condition in the MSSM. These corrections are sizable and important in the relevant regions of parameter space and furthermore give rise to solutions of the alignment condition that are not present in the approximate one-loop description. We provide a comprehensive numerical comparison of the alignment condition obtained in the approximate one-loop and two-loop approximations, and discuss its implications for phenomenologically viable regions of the MSSM parameter space. (orig.)

  11. Testing SUSY at the LHC: Electroweak and Dark matter fine tuning at two-loop order

    CERN Document Server

    Cassel, S; Ross, G G

    2010-01-01

    In the framework of the Constrained Minimal Supersymmetric Standard Model (CMSSM) we evaluate the electroweak fine tuning measure that provides a quantitative test of supersymmetry as a solution to the hierarchy problem. Taking account of current experimental constraints we compute the fine tuning at two-loop order and determine the limits on the CMSSM parameter space and the measurements at the LHC most relevant in covering it. Without imposing the LEPII bound on the Higgs mass, it is shown that the fine tuning computed at two-loop has a minimum $\\Delta=8.8$ corresponding to a Higgs mass $m_h=114\\pm 2$ GeV. Adding the constraint that the SUSY dark matter relic density should be within present bounds we find $\\Delta=15$ corresponding to $m_h=114.7\\pm 2$ GeV and this rises to $\\Delta=17.8$ ($m_h=115.9\\pm 2$ GeV) for SUSY dark matter abundance within 3$\\sigma$ of the WMAP constraint. We extend the analysis to include the contribution of dark matter fine tuning. In this case the overall fine tuning and Higgs mas...

  12. Two-loop scale-invariant scalar potential and quantum effective operators

    CERN Document Server

    Ghilencea, D.M.

    2016-11-29

    Spontaneous breaking of quantum scale invariance may provide a solution to the hierarchy and cosmological constant problems. In a scale-invariant regularization, we compute the two-loop potential of a higgs-like scalar $\\phi$ in theories in which scale symmetry is broken only spontaneously by the dilaton ($\\sigma$). Its vev $\\langle\\sigma\\rangle$ generates the DR subtraction scale ($\\mu\\sim\\langle\\sigma\\rangle$), which avoids the explicit scale symmetry breaking by traditional regularizations (where $\\mu$=fixed scale). The two-loop potential contains effective operators of non-polynomial nature as well as new corrections, beyond those obtained with explicit breaking ($\\mu$=fixed scale). These operators have the form: $\\phi^6/\\sigma^2$, $\\phi^8/\\sigma^4$, etc, which generate an infinite series of higher dimensional polynomial operators upon expansion about $\\langle\\sigma\\rangle\\gg \\langle\\phi\\rangle$, where such hierarchy is arranged by {\\it one} initial, classical tuning. These operators emerge at the quantum...

  13. Infrared divergences and harmonic anomalies in the two-loop superstring effective action

    CERN Document Server

    Pioline, Boris

    2015-01-01

    We analyze the pertubative contributions to the $D^4 R^4$ and $D^6 R^4$ couplings in the low-energy effective action of type II string theory compactified on a torus $T^d$, with particular emphasis on two-loop corrections. In general, it is necessary to introduce an infrared cut-off $\\Lambda$ to separate local interactions from non-local effects due to the exchange of massless states. We identify the degenerations of the genus-two Riemann surface which are responsible for power-like dependence on $\\Lambda$, and give an explicit prescription for extracting the $\\Lambda$-independent effective couplings. These renormalized couplings are then shown to be eigenmodes of the Laplace operator with respect to the torus moduli, up to computable anomalous source terms arising in the presence of logarithmic divergences, in precise agreement with predictions from U-duality. Our results for the two-loop $D^6 R^4$ contribution also probe essential properties of the Kawazumi-Zhang invariant

  14. Infrared divergences and harmonic anomalies in the two-loop superstring effective action

    Energy Technology Data Exchange (ETDEWEB)

    Pioline, Boris [CERN PH-TH,Case C01600, CERN, CH-1211 Geneva 23 (Switzerland); Sorbonne Universités,UPMC Université Paris 6, UMR 7589, F-75005 Paris (France); Laboratoire de Physique Théorique et Hautes Energies, CNRS UMR 7589,Université Pierre et Marie Curie, 4 place Jussieu, 75252 Paris cedex 05 (France); Russo, Rodolfo [Centre for Research in String Theory, School of Physics and Astronomy,Queen Mary University of London, Mile End Road, London, E1 4NS (United Kingdom)

    2015-12-16

    We analyze the pertubative contributions to the D{sup 4}R{sup 4} and D{sup 6}R{sup 4} couplings in the low-energy effective action of type II string theory compactified on a torus T{sup d}, with particular emphasis on two-loop corrections. In general, it is necessary to introduce an infrared cut-off Λ to separate local interactions from non-local effects due to the exchange of massless states. We identify the degenerations of the genus-two Riemann surface which are responsible for power-like dependence on Λ, and give an explicit prescription for extracting the Λ-independent effective couplings. These renormalized couplings are then shown to be eigenmodes of the Laplace operator with respect to the torus moduli, up to computable anomalous source terms arising in the presence of logarithmic divergences, in precise agreement with predictions from U-duality. Our results for the two-loop D{sup 6}R{sup 4} contribution also probe essential properties of the Kawazumi-Zhang invariant.

  15. Dynamical symmetry breaking of λφ4 theory in the two loop effective potential

    International Nuclear Information System (INIS)

    Yang Jifeng; Ruan Jianhong

    2002-01-01

    The two loop effective potential of massless λφ 4 theory is presented in several regularization and renormalization prescriptions and the dynamical symmetry breaking solution is obtained in the strong-coupling situation in several prescriptions except the Coleman-Weinberg prescription. The beta function in the broken phase becomes negative and the UV fixed point turns out to be a strong-coupling one, and its numeric value varies with the renormalization prescriptions, a detail which is different from the asymptotic-free solution in the one loop case. The symmetry-breaking phase is shown to be an entirely strong-coupling phase. The reason for the relevance of the renormalization prescriptions is shown to be due to the nonperturbative nature of the effective potential. We also reanalyze the two loop effective potential by adopting a differential equation approach based on the understanding that all the quantum field theories are ill-defined formulations of the 'low-energy' effective theories of a complete underlying theory. The relevance of the prescriptions of fixing the local ambiguities to physical properties such as symmetry breaking is further emphasized. We also tentatively propose a rescaling insensitivity argument for fixing the quadratic ambiguities. Some detailed properties of the strongly coupled broken phase and related issues are discussed

  16. Two-loop corrections to the ρ parameter in Two-Higgs-Doublet models

    Energy Technology Data Exchange (ETDEWEB)

    Hessenberger, Stephan; Hollik, Wolfgang [Max-Planck-Institut fuer Physik (Werner-Heisenberg-Institut), Muenchen (Germany)

    2017-03-15

    Models with two scalar doublets are among the simplest extensions of the Standard Model which fulfill the relation ρ = 1 at lowest order for the ρ parameter as favored by experimental data for electroweak observables allowing only small deviations from unity. Such small deviations Δρ originate exclusively from quantum effects with special sensitivity to mass splittings between different isospin components of fermions and scalars. In this paper the dominant two-loop electroweak corrections to Δρ are calculated in the CP-conserving THDM, resulting from the top-Yukawa coupling and the self-couplings of the Higgs bosons in the gauge-less limit. The on-shell renormalization scheme is applied. With the assumption that one of the CP-even neutral scalars represents the scalar boson observed by the LHC experiments, with standard properties, the two-loop non-standard contributions in Δρ can be separated from the standard ones. These contributions are of particular interest since they increase with mass splittings between non-standard Higgs bosons and can be additionally enhanced by tanβ and λ{sub 5}, an additional free coefficient of the Higgs potential, and can thus modify the one-loop result substantially. Numerical results are given for the dependence on the various non-standard parameters, and the influence on the calculation of electroweak precision observables is discussed. (orig.)

  17. Two-loop disorder effects on the nematic quantum criticality in d-wave superconductors

    International Nuclear Information System (INIS)

    Wang, Jing

    2015-01-01

    The gapless nodal fermions exhibit non-Fermi liquid behaviors at the nematic quantum critical point that is supposed to exist in some d-wave cuprate superconductors. This non-Fermi liquid state may be turned into a disorder-dominated diffusive metal if the fermions also couple to a disordered potential that generates a relevant perturbation in the sense of renormalization group theory. It is therefore necessary to examine whether a specific disorder is relevant or not. We study the interplay between critical nematic fluctuation and random chemical potential by performing renormalization group analysis. The parameter that characterizes the strength of random chemical potential is marginal at the one-loop level, but becomes marginally relevant after including the two-loop corrections. Thus even weak random chemical potential leads to diffusive motion of nodal fermions and the significantly critical behaviors of physical implications, since the strength flows eventually to large values at low energies. - Highlights: • The gapless nodal fermions exhibit non-Fermi liquid behaviors at the nematic QCP. • The strength of random chemical potential is marginal at the one-loop level. • The strength becomes marginally relevant after including the two-loop corrections. • The diffusive metallic state is induced by the marginally relevant disorder. • The behaviors of some physical observables are presented at the nematic QCP

  18. Westinghouse waste simulation and optimization software tool

    International Nuclear Information System (INIS)

    Mennicken, Kim; Aign, Jorg

    2013-01-01

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

  19. Westinghouse waste simulation and optimization software tool

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  20. Benchmarking Computational Fluid Dynamics for Application to PWR Fuel

    International Nuclear Information System (INIS)

    Smith, L.D. III; Conner, M.E.; Liu, B.; Dzodzo, B.; Paramonov, D.V.; Beasley, D.E.; Langford, H.M.; Holloway, M.V.

    2002-01-01

    The present study demonstrates a process used to develop confidence in Computational Fluid Dynamics (CFD) as a tool to investigate flow and temperature distributions in a PWR fuel bundle. The velocity and temperature fields produced by a mixing spacer grid of a PWR fuel assembly are quite complex. Before using CFD to evaluate these flow fields, a rigorous benchmarking effort should be performed to ensure that reasonable results are obtained. Westinghouse has developed a method to quantitatively benchmark CFD tools against data at conditions representative of the PWR. Several measurements in a 5 x 5 rod bundle were performed. Lateral flow-field testing employed visualization techniques and Particle Image Velocimetry (PIV). Heat transfer testing involved measurements of the single-phase heat transfer coefficient downstream of the spacer grid. These test results were used to compare with CFD predictions. Among the parameters optimized in the CFD models based on this comparison with data include computational mesh, turbulence model, and boundary conditions. As an outcome of this effort, a methodology was developed for CFD modeling that provides confidence in the numerical results. (authors)

  1. 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

  2. Multi-loop PWR modeling and hardware-in-the-loop testing using ACSL

    International Nuclear Information System (INIS)

    Thomas, V.M.; Heibel, M.D.; Catullo, W.J.

    1989-01-01

    Westinghouse has developed an Advanced Digital Feedwater Control System (ADFCS) which is aimed at reducing feedwater related reactor trips through improved control performance for pressurized water reactor (PWR) power plants. To support control system setpoint studies and functional design efforts for the ADFCS, an ACSL based model of the nuclear steam supply system (NSSS) of a Westinghouse (PWR) was generated. Use of this plant model has been extended from system design to system testing through integration of the model into a Hardware-in-Loop test environment for the ADFCS. This integration includes appropriate interfacing between a Gould SEL 32/87 computer, upon which the plant model executes in real time, and the Westinghouse Distributed Processing family (WDPF) test hardware. A development program has been undertaken to expand the existing ACSL model to include capability to explicitly model multiple plant loops, steam generators, and corresponding feedwater systems. Furthermore, the program expands the ADFCS Hardware-in-Loop testing to include the multi-loop plant model. This paper provides an overview of the testing approach utilized for the ADFCS with focus on the role of Hardware-in-Loop testing. Background on the plant model, methodology and test environment is also provided. Finally, an overview is presented of the program to expand the model and associated Hardware-in-Loop test environment to handle multiple loops

  3. 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

  4. Two-loop integrals for CP-even heavy quarkonium production and decays: elliptic sectors

    Science.gov (United States)

    Chen, Long-Bin; Jiang, Jun; Qiao, Cong-Feng

    2018-04-01

    By employing the differential equations, we compute analytically the elliptic sectors of two-loop master integrals appearing in the NNLO QCD corrections to CP-even heavy quarkonium exclusive production and decays, which turns out to be the last and toughest part in the relevant calculation. The integrals are found can be expressed as Goncharov polylogarithms and iterative integrals over elliptic functions. The master integrals may be applied to some other NNLO QCD calculations about heavy quarkonium exclusive production, like {γ}^{\\ast}γ \\to Q\\overline{Q} , {e}+{e}-\\to γ +Q\\overline{Q} , and H/{Z}^0\\to γ +Q\\overline{Q} , heavy quarkonium exclusive decays, and also the CP-even heavy quarkonium inclusive production and decays.

  5. The Higgs Mass in the MSSM at two-loop order beyond minimal flavour violation

    CERN Document Server

    Goodsell, Mark D; Staub, Florian

    2016-01-01

    Soft supersymmetry-breaking terms provide a wealth of new potential sources of flavour violation, which lead to very tight constraints from precision experiments. This has posed a challenge to construct flavour models to both explain the structure of the Standard Model Yukawa couplings and how their consequent predictions for patterns in the soft supersymmetry-breaking terms do not violate these constraints. While such models have been studied in great detail, the impact of flavour violating soft terms on the Higgs mass at the two-loop level has been assumed to be small or negligible. In this letter, we show that large flavour violation in the up-squark sector can give a positive or negative shift to the SM-like Higgs of several GeV, without being in conflict with any other observation. We investigate in which regions of the parameter space these effects can be expected.

  6. A comprehensive coordinate space renormalization of quantum electrodynamics to two-loop order

    International Nuclear Information System (INIS)

    Haagensen, P.E.; Latorre, J.I.

    1993-01-01

    We develop a coordinate space renormalization of massless quantum electrodynamics using the powerful method of differential renormalization. Bare one-loop amplitudes are finite at non-coincident external points, but do not accept a Fourier transform into momentum space. The method provides a systematic procedure to obtain one-loop renormalized amplitudes with finite Fourier transforms in strictly four dimensions without the appearance of integrals or the use of a regulator. Higher loops are solved similarly by renormalizing from the inner singularities outwards to the global one. We compute all one- and two-loop 1PI diagrams, run renormalization group equations on them. and check Ward identities. The method furthermore allows us to discern a particular pattern of renormalization under which certain amplitudes are seen not to contain higher-loop leading logarithms. We finally present the computation of the chiral triangle showing that differential renormalization emerges as a natural scheme to tackle γ 5 problems

  7. Experimental Study of Flexible Plate Vibration Control by Using Two-Loop Sliding Mode Control Strategy

    Science.gov (United States)

    Yang, Jingyu; Lin, Jiahui; Liu, Yuejun; Yang, Kang; Zhou, Lanwei; Chen, Guoping

    2017-08-01

    It is well known that intelligent control theory has been used in many research fields, novel modeling method (DROMM) is used for flexible rectangular active vibration control, and then the validity of new model is confirmed by comparing finite element model with new model. In this paper, taking advantage of the dynamics of flexible rectangular plate, a two-loop sliding mode (TSM) MIMO approach is introduced for designing multiple-input multiple-output continuous vibration control system, which can overcome uncertainties, disturbances or unstable dynamics. An illustrative example is given in order to show the feasibility of the method. Numerical simulations and experiment confirm the effectiveness of the proposed TSM MIMO controller.

  8. Small-threshold behaviour of two-loop self-energy diagrams: two-particle thresholds

    International Nuclear Information System (INIS)

    Berends, F.A.; Davydychev, A.I.; Moskovskij Gosudarstvennyj Univ., Moscow; Smirnov, V.A.; Moskovskij Gosudarstvennyj Univ., Moscow

    1996-01-01

    The behaviour of two-loop two-point diagrams at non-zero thresholds corresponding to two-particle cuts is analyzed. The masses involved in a cut and the external momentum are assumed to be small as compared to some of the other masses of the diagram. By employing general formulae of asymptotic expansions of Feynman diagrams in momenta and masses, we construct an algorithm to derive analytic approximations to the diagrams. In such a way, we calculate several first coefficients of the expansion. Since no conditions on relative values of the small masses and the external momentum are imposed, the threshold irregularities are described analytically. Numerical examples, using diagrams occurring in the standard model, illustrate the convergence of the expansion below the first large threshold. (orig.)

  9. TLTA/6431, Two-Loop-Test-Apparatus, BWR/6 Simulator, Small-Break LOCA

    International Nuclear Information System (INIS)

    1992-01-01

    1 - Description of test facility: The Two-Loop-Test-Apparatus (TLTA) is a 1:624 volume scaled BWR/6 simulator. It was the predecessor of the better-scaled FIST facility. The facility is capable of full BWR system pressure and has a simulated core with a full size 8 x 8, full power single bundle of indirect electrically heated rods. All major BWR systems are simulated including lower plenum, guide tube, core region (bundle and bypass), upper plenum, steam separator, steam dome, annular downcomer, recirculation loops and ECC injection systems. The fundamental scaling consideration was to achieve real-time response. A number of the scaling compromises present in TLTA were corrected in the FIST configuration. These compromises include a number of regional volumes and component elevations. 2 - Description of test: 64.45 sqcm small break LOCA with activation of the full emergency core cooling system, but without activation of the automatic decompression system

  10. Probing the desert by the two-loop renormalization-group equations

    International Nuclear Information System (INIS)

    Tanimoto, M.; Suetake, Y.; Senba, K.

    1987-01-01

    We have reexamined the study of probing the desert with fermion masses, presented by Bagger, Dimopoulos, and Masso, by using the two-loop renormalization-group equations in the framework of the SU(3) x SU(2) x U(1) model with three generations and one Higgs doublet. The blow-up energy scale of the Yukawa coupling is found to be dependent on the Higgs quartic coupling λ. If the Yukawa coupling blows up between the electroweak scale M/sub W/ and the grand unified scale M/sub X/, the Higgs potential is destabilized for small values of λ at the electroweak scale M/sub W/, and becomes strongly coupled for large values of λ at M/sub W/. It is found that the Higgs-scalar mass as well as the fermion masses are important to probe the desert

  11. Evolution of the pion wave function in the scalar /phi/63 model: two-loop calculation

    International Nuclear Information System (INIS)

    Mikhailov, S.V.; Radyushkin, A.V.

    1986-01-01

    The authors study the structure of the contributions that violate the multiplicative renormalizability of the conformal operators in the model based on the /phi/ 6 3 theory in space-time of six dimensions. This theory has a number of features in common with QCD in four dimensions. The basic propositions are presented and the key elements of the calculation are demonstrated. The connection between the kernels for exclusive and inclusive processes are discused and the structure of the two-loop evolution kernel V(x,y) and the solution of the evolution equation are discussed. Main conclusions are formulated and the results of the calculations for concrete diagrams are deferred to in Appendix A. Formulas for the transition from the exclusive to the inclusive kernels are presented in Appendix B

  12. Dominant two-loop corrections to the MSSM finite temperature effective potential

    International Nuclear Information System (INIS)

    Espinosa, J.R.

    1996-04-01

    We show that two-loop corrections to the finite temperature effective potential in the MSSM can have a dramatic effect on the strength of the electroweak phase transition, making it more strongly first order. The change in the order parameter v/Tc can be as large as 75% of the one-loop daisy improved result. This effect can be decisive to widen the region in parameter space where erasure of the created baryons by sphaleron processes after the transition is suppressed and hence, where electroweak baryogenesis might be successful. We find an allowed region with tan β< or∼4.5 and a Higgs boson with standard couplings and mass below 80 GeV within the reach of LEP II. (orig.)

  13. Complete two-loop effective potential approximation to the lightest Higgs scalar boson mass in supersymmetry

    International Nuclear Information System (INIS)

    Martin, Stephen P.

    2003-01-01

    I present a method for accurately calculating the pole mass of the lightest Higgs scalar boson in supersymmetric extensions of the standard model, using a mass-independent renormalization scheme. The Higgs scalar self-energies are approximated by supplementing the exact one-loop results with the second derivatives of the complete two-loop effective potential in Landau gauge. I discuss the dependence of this approximation on the choice of renormalization scale, and note the existence of particularly poor choices, which fortunately can be easily identified and avoided. For typical input parameters, the variation in the calculated Higgs boson mass over a wide range of renormalization scales is found to be of the order of a few hundred MeV or less, and is significantly improved over previous approximations

  14. PWR core design calculations

    International Nuclear Information System (INIS)

    Trkov, A.; Ravnik, M.; Zeleznik, N.

    1992-01-01

    Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [sl

  15. Next generation PWR

    International Nuclear Information System (INIS)

    Tanaka, Toshihiko; Fukuda, Toshihiko; Usui, Shuji

    2001-01-01

    Development of LWR for power generation in Japan has been intended to upgrade its reliability, safety, operability, maintenance and economy as well as to increase its capacity in order, since nuclear power generation for commercial use was begun on 1970, to steadily increase its generation power. And, in Japan, ABWR (advanced BWR) of the most promising LWR in the world, was already used actually and APWR (advanced PWR) with the largest output in the world is also at a step of its actual use. And, development of the APWR in Japan was begun on 1980s, and is at a step of plan on construction of its first machine at early of this century. However, by large change of social affairs, economy of nuclear power generation is extremely required, to be positioned at an APWR improved development reactor promoted by collaboration of five PWR generation companies and the Mitsubishi Electric Co., Ltd. Therefore, on its development, investigation on effect of change in social affairs on nuclear power stations was at first carried out, to establish a design requirement for the next generation PWR. Here were described on outline, reactor core design, safety concept, and safety evaluation of APWR+ and development of an innovative PWR. (G.K.)

  16. 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

  17. 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)

  18. Construction of PWR nuclear cross sections for transient calculations. Test of the ANTI program against TWODIM

    International Nuclear Information System (INIS)

    Thorlaksen, B.

    1981-05-01

    Nuclear cross sections for fuel assemblies of the more recent Westinghouse designs, representing two different PWR reactor cores, are calculated as functions of average fuel temperature, moderator density, and moderator poison concentration. The cross-section functions are verified by referring to Westinghouse power-shape calculations and other analysis. Computations on the side reflector resulted in significantly higher albedo values than used previously for BWR's in similar nodal codes. This led to an investigation of the influence of the internodal coupling coefficients on the power shape. It is concluded that the calculated power shape is strongly dependent, on the choise of coupling coefficients. However, it is shown that ''the correct'' set of coupling coefficients depends mostly on the nodal configuration, and that it is fairly independent of the power condition. (author)

  19. 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.

  20. 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

  1. 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)

  2. 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

  3. Evidence for two-loop interaction from IRIS and SDO observations of penumbral brightenings

    Science.gov (United States)

    Alissandrakis, C. E.; Koukras, A.; Patsourakos, S.; Nindos, A.

    2017-07-01

    Aims: We investigate small scale energy release events which can provide clues on the heating mechanism of the solar corona. Methods: We analyzed spectral and imaging data from the Interface Region Imaging Spectrograph (IRIS), images from the Atmospheric Imaging Assembly (AIA) aboard the Solar Dynamics Observatoty (SDO), and magnetograms from the Helioseismic and Magnetic Imager (HMI) aboard SDO. Results: We report observations of small flaring loops in the penumbra of a large sunspot on July 19, 2013. Our main event consisted of a loop spanning 15'', from the umbral-penumbral boundary to an opposite polarity region outside the penumbra. It lasted approximately 10 min with a two minute impulsive peak and was observed in all AIA/SDO channels, while the IRIS slit was located near its penumbral footpoint. Mass motions with an apparent velocity of 100 km s-1 were detected beyond the brightening, starting in the rise phase of the impulsive peak; these were apparently associated with a higher-lying loop. We interpret these motions in terms of two-loop interaction. IRIS spectra in both the C II and Si iv lines showed very extended wings, up to about 400 km s-1, first in the blue (upflows) and subsequently in the red wing. In addition to the strong lines, emission was detected in the weak lines of Cl I, O I and C I, as well as in the Mg II triplet lines. Absorption features in the profiles of the C II doublet, the Si iv doublet and the Mg II h and k lines indicate the existence of material with a lower source function between the brightening and the observer. We attribute this absorption to the higher loop and this adds further credibility to the two-loop interaction hypothesis. Tilts were detected in the absorption spectra, as well as in the spectra of Cl I, O I, and C I lines, possibly indicating rotational motions from the untwisting of magnetic flux tubes. Conclusions: We conclude that the absorption features in the C II, Si iv and Mg II profiles originate in a higher

  4. Experimental studies of PWR primary piping under loca

    International Nuclear Information System (INIS)

    Caumette, Pierre; Garcia, J.L.

    1980-07-01

    The experimental program performed on AQUITAINE II facility is directed to study the mechanical behavior of primary PWR pipes and the forces exerted on the neighbouring structures as a consequence of a breach opening. It has been developed in the form of a quadripartite agreement between the Commissariat a l'Energie Atomique, Framatome, Electricite de France and Westinghouse. Some forty tests have been carried out with different pipe configurations (straight tube, elbow, S- or U-shaped tube) and different break types (single or double guillotine). The following aspects are investigated: - the dynamic behavior of the pipe and in particular the formation of a plastic hinge at the restraint; - the impact function of a pipe or an energy-absorbing bumper; - the lateral stability of both ends of a pipe, after a double-guillotine break [fr

  5. Failure probability of PWR reactor coolant loop piping

    International Nuclear Information System (INIS)

    Lo, T.; Woo, H.H.; Holman, G.S.; Chou, C.K.

    1984-02-01

    This paper describes the results of assessments performed on the PWR coolant loop piping of Westinghouse and Combustion Engineering plants. For direct double-ended guillotine break (DEGB), consideration was given to crack existence probability, initial crack size distribution, hydrostatic proof test, preservice inspection, leak detection probability, crack growth characteristics, and failure criteria based on the net section stress failure and tearing modulus stability concept. For indirect DEGB, fragilities of major component supports were estimated. The system level fragility was then calculated based on the Boolean expression involving these fragilities. Indirect DEGB due to seismic effects was calculated by convolving the system level fragility and the seismic hazard curve. The results indicate that the probability of occurrence of both direct and indirect DEGB is extremely small, thus, postulation of DEGB in design should be eliminated and replaced by more realistic criteria

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

    International Nuclear Information System (INIS)

    Sechensky, B.

    2005-01-01

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

  7. 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

  8. Scaling studies - PWR

    International Nuclear Information System (INIS)

    Sonneck, G.

    1983-05-01

    A RELAP 4/MOD 6 study was made based on the blowdown phase of the intermediate break experiment LOFT L5-1. The method was to set up a base model and to vary parametrically some areas where it is known or suspected that LOFT differs from a commercial PWR. The aim was not to simulate LOFT or a PWR exactly but to understand the influence of the following parameters on the thermohydraulic behaviour of the system and the clad temperature: stored heat in the downcomer (LOFT has rather large filler blocks in this part of the pressure vessel); bypass between downcomer and upper plenum; and core length. The results show that LOFT is prototypical for all calculated blowdowns. As the clad temperatures decrease with decreasing stored energy in the downcomer, increased bypass and increased core length, LOFT results seem to be realistic as long as realistic bypass sizes are considered; they are conservative in the two other areas. (author)

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  10. Pole Mass of the W Boson at Two-Loop Order in the Pure $\\overline {MS}$ Scheme

    Energy Technology Data Exchange (ETDEWEB)

    Martin, Stephen P. [Northern Illinois U.

    2015-06-03

    I provide a calculation at full two-loop order of the complex pole squared mass of the W boson in the Standard Model in the pure MS¯ renormalization scheme, with Goldstone boson mass effects resummed. This approach is an alternative to earlier ones that use on-shell or hybrid renormalization schemes. The renormalization scale dependence of the real and imaginary parts of the resulting pole mass is studied. Both deviate by about ±4  MeV from their median values as the renormalization scale is varied from 50 to 200 GeV, but the theory error is likely larger. A surprising feature of this scheme is that the two-loop QCD correction has a larger scale dependence, but a smaller magnitude, than the two-loop non-QCD correction, unless the renormalization scale is chosen very far from the top-quark mass.

  11. Plutonium recycling in PWR

    International Nuclear Information System (INIS)

    Youinou, G.; Girieud, R.; Guigon, B.

    2000-01-01

    Two concepts of 100% MOX PWR cores are presented. They are designed such as to minimize the consequences of the introduction of Pu on the core control. The first one has a high moderation ratio and the second one utilizes an enriched uranium support. The important design parameters as well as their capabilities to multi recycle Pu are discussed. We conclude with the potential interest of the two concepts. (author)

  12. The integrated PWR

    International Nuclear Information System (INIS)

    Gautier, G.M.

    2002-01-01

    This document presents the integrated reactors concepts by a presentation of four reactors: PIUS, SIR, IRIS and CAREM. The core conception, the operating, the safety, the economical aspects and the possible users are detailed. From the performance of the classical integrated PWR, the necessity of new innovative fuels utilization, the research of a simplified design to make easier the safety and the KWh cost decrease, a new integrated reactor is presented: SCAR 600. (A.L.B.)

  13. 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)

  14. Experimental tests for the Babu-Zee two-loop model of Majorana neutrino masses

    International Nuclear Information System (INIS)

    Sierra, Diego Aristizabal; Hirsch, Martin

    2006-01-01

    The smallness of the observed neutrino masses might have a radiative origin. Here we revisit a specific two-loop model of neutrino mass, independently proposed by Babu and Zee. We point out that current constraints from neutrino data can be used to derive strict lower limits on the branching ratio of flavour changing charged lepton decays, such as μ→eγ. Non-observation of Br(μ→eγ) at the level of 10 -13 would rule out singly charged scalar masses smaller than 590 GeV (5.04 TeV) in case of normal (inverse) neutrino mass hierarchy. Conversely, decay branching ratios of the non-standard scalars of the model can be fixed by the measured neutrino angles (and mass scale). Thus, if the scalars of the model are light enough to be produced at the LHC or ILC, measuring their decay properties would serve as a direct test of the model as the origin of neutrino masses

  15. Experimental tests for the Babu-Zee two-loop model of Majorana neutrino masses

    International Nuclear Information System (INIS)

    Aristizabal, D.

    2006-01-01

    Abstract: The smallness of the observed neutrino masses might have a radiative origin. Here we revisit a specific two-loop model of neutrino mass, independently proposed by Babu and Zee. We point out that current constraints from neutrino data can be used to derive strict lower limits on the branching ratio of flavour changing charged lepton decays, such as μ → e γ. Non-observation of Br(μ → e γ) at the level of 10 -13 would rule out singly charged scalar masses smaller than 590 GeV (5.04 TeV) in case of normal (inverse) neutrino mass hierarchy. Conversely, decay branching ratios of the non-standard scalars of the model can be fixed by the measured neutrino angles (and mass scale). Thus, if the scalars of the model are light enough to be produced at the LHC or ILC, measuring their decay properties would serve as a direct test of the model as the origin of neutrino masses. (author)

  16. Systematic classification of two-loop realizations of the Weinberg operator

    Energy Technology Data Exchange (ETDEWEB)

    Sierra, D. Aristizabal; Degee, A. [IFPA, Dep. AGO, Universite de Liege,Bat B5, Sart Tilman B-4000 Liege 1 (Belgium); Dorame, L.; Hirsch, M. [AHEP Group, Instituto de Fisica Corpuscular-C.S.I.C./Universitat de Valencia,Edificio Institutos de Paterna, Apt 22085, E-46071 Valencia (Spain)

    2015-03-09

    We systematically analyze the d=5 Weinberg operator at 2-loop order. Using a diagrammatic approach, we identify two different interesting categories of neutrino mass models: (i) Genuine 2-loop models for which both, tree-level and 1-loop contributions, are guaranteed to be absent. And (ii) finite 2-loop diagrams, which correspond to the 1-loop generation of some particular vertex appearing in a given 1-loop neutrino mass model, thus being effectively 2-loop. From the large list of all possible 2-loop diagrams, the vast majority are infinite corrections to lower order neutrino mass models and only a moderately small number of diagrams fall into these two interesting classes. Moreover, all diagrams in class (i) are just variations of three basic diagrams, with examples discussed in the literature before. Similarly, we also show that class (ii) diagrams consists of only variations of these three plus two more basic diagrams. Finally, we show how our results can be consistently and readily used in order to construct two-loop neutrino mass models.

  17. Avoiding the Goldstone Boson Catastrophe in general renormalisable field theories at two loops

    Energy Technology Data Exchange (ETDEWEB)

    Braathen, Johannes; Goodsell, Mark D. [LPTHE, UPMC University Paris 06, Sorbonne Universités,4 Place Jussieu, F-75252 Paris (France); LPTHE, CNRS,4 Place Jussieu, F-75252 Paris (France)

    2016-12-14

    We show how the infra-red divergences associated to Goldstone bosons in the minimum condition of the two-loop Landau-gauge effective potential can be avoided in general field theories. This extends the resummation formalism recently developed for the Standard Model and the MSSM, and we give compact, infra-red finite expressions in closed form for the tadpole equations. We also show that the results at this loop order are equivalent to (and are most easily obtained by) imposing an “on-shell” condition for the Goldstone bosons. Moreover, we extend the approach to show how the infra-red divergences in the calculation of the masses of neutral scalars (such as the Higgs boson) can be eliminated. For the mass computation, we specialise to the gaugeless limit and extend the effective potential computation to allow the masses to be determined without needing to solve differential equations for the loop functions — opening the door to fast, infra-red safe determinations of the Higgs mass in general theories.

  18. Two-loop conformal generators for leading-twist operators in QCD

    International Nuclear Information System (INIS)

    Braun, V.M.; Strohmaier, M.; Manashov, A.N.; Hamburg Univ.; Moch, S.

    2016-01-01

    QCD evolution equations in minimal subtraction schemes have a hidden symmetry: One can construct three operators that commute with the evolution kernel and form an SL(2) algebra, i.e. they satisfy (exactly) the SL(2) commutation relations. In this paper we find explicit expressions for these operators to two-loop accuracy going over to QCD in non-integer d=4-2ε space-time dimensions at the intermediate stage. In this way conformal symmetry of QCD is restored on quantum level at the specially chosen (critical) value of the coupling, and at the same time the theory is regularized allowing one to use the standard renormalization procedure for the relevant Feynman diagrams. Quantum corrections to conformal generators in d=4-2ε effectively correspond to the conformal symmetry breaking in the physical theory in four dimensions and the SL(2) commutation relations lead to nontrivial constraints on the renormalization group equations for composite operators. This approach is valid to all orders in perturbation theory and the result includes automatically all terms that can be identified as due to a nonvanishing QCD β-function (in the physical theory in four dimensions). Our result can be used to derive three-loop evolution equations for flavor-nonsinglet quark-antiquark operators including mixing with the operators containing total derivatives. These equations govern, e.g., the scale dependence of generalized hadron parton distributions and light-cone meson distribution amplitudes.

  19. Two-loop beam and soft functions for rapidity-dependent jet vetoes

    Energy Technology Data Exchange (ETDEWEB)

    Gangal, Shireen [Theory Group, Deutsches Elektronen-Synchrotron (DESY),Notkestraße 85, D-22607 Hamburg (Germany); Gaunt, Jonathan R. [Nikhef Theory Group and VU University Amsterdam,De Boelelaan 1081, NL-1081 HV Amsterdam (Netherlands); Stahlhofen, Maximilian [PRISMA Cluster of Excellence, Institute of Physics, Johannes Gutenberg University,Staudingerweg 7, D-55128 Mainz (Germany); Tackmann, Frank J. [Theory Group, Deutsches Elektronen-Synchrotron (DESY),Notkestraße 85, D-22607 Hamburg (Germany)

    2017-02-06

    Jet vetoes play an important role in many analyses at the LHC. Traditionally, jet vetoes have been imposed using a restriction on the transverse momentum p{sub Tj} of jets. Alternatively, one can also consider jet observables for which p{sub Tj} is weighted by a smooth function of the jet rapidity y{sub j} that vanishes as |y{sub j}|→∞. Such observables are useful as they provide a natural way to impose a tight veto on central jets but a looser one at forward rapidities. We consider two such rapidity-dependent jet veto observables, T{sub Bj} and T{sub Cj}, and compute the required beam and dijet soft functions for the jet-vetoed color-singlet production cross section at two loops. At this order, clustering effects from the jet algorithm become important. The dominant contributions are computed fully analytically while corrections that are subleading in the limit of small jet radii are expressed in terms of finite numerical integrals. Our results enable the full NNLL{sup ′} resummation and are an important step towards N{sup 3}LL resummation for cross sections with a T{sub Bj} or T{sub Cj} jet veto.

  20. Reactor control system. PWR

    International Nuclear Information System (INIS)

    2009-01-01

    At present, 23 units of PWR type reactors have been operated in Japan since the start of Mihama Unit 1 operation in 1970 and various improvements have been made to upgrade operability of power stations as well as reliability and safety of power plants. As the share of nuclear power increases, further improvements of operating performance such as load following capability will be requested for power stations with more reliable and safer operation. This article outlined the reactor control system of PWR type reactors and described the control performance of power plants realized with those systems. The PWR control system is characterized that the turbine power is automatic or manually controlled with request of the electric power system and then the nuclear power is followingly controlled with the change of core reactivity. The system mainly consists of reactor automatic control system (control rod control system), pressurizer pressure control system, pressurizer water level control system, steam generator water level control system and turbine bypass control system. (T. Tanaka)

  1. AGR v PWR

    International Nuclear Information System (INIS)

    Green, D.

    1986-01-01

    When the Central Electricity Generating Board (CEGB) invited tenders and placed a contract for the Advanced Gas Cooled Reactor (AGR) at Dungeness B in 1965 -preferring it to the Pressurised Water Reactor (PWR) -the AGR was lamentably ill developed. The effects of the decision were widely felt, for it took the British nuclear industry off the light water reactor highway of world reactor business and up and idiosyncratic private highway of its own, excluding it altogether from any material export business in the two decades which followed. Yet although the UK may have made wrong decisions in rejecting the PWR in 1965, that does not mean that it can necessarily now either correct them, or redeem their consequence, by reversing the choice in 1985. In the 20 years since 1965 the whole world economic and energy picture has been transformed and the national picture with it. Picking up the PWR now could prove as big a disaster as rejecting it may have been in 1965. (author)

  2. Water chemistry in PWR

    International Nuclear Information System (INIS)

    Abe, Kenji

    1987-01-01

    This article outlines major features and basic concept of the secondary system of PWR's and water properties control measures adopted in recent PWR plants. The secondary system of a PWR consists of a condenser cooling pipe (aluminum-brass, titanium, or stainless steel), low-pressure make-up water heating pipe (aluminum-brass or stainless steel), high-ressure make-up water heating pipe (cupro-nickel or stainless steel), steam generator heat-transfer pipe (Inconel 600 or 690), and bleed/drain pipe (carbon steel, low alloy steel or stainless steel). Other major pipes and equipment are made of carbon steel or stainless steel. Major troubles likely to be caused by water in the secondary system include reduction in wall thickness of the heat-transfer pipe, stress corrosion cracking in the heat-transfer pipe, and denting. All of these are caused by local corrosion due to concentration of purities contained in water. For controlling the water properties in the secondary system, it is necessary to prevent impurities from entering the system, to remove impurities and corrosion products from the system, and to prevent corrosion of apparatus making up the system. Measures widely adopted for controlling the formation of IGA include the addition of boric acid for decreasing the concentration of free alkali and high hydrazine operation for providing a highly reducing atmospere. (Nogami, K.)

  3. (Quasi)Elastic Electron-Muon Large-Angle Scattering to a Two-Loop Approximation: Vertex Contributions

    CERN Document Server

    Bytev, V V; Shaikhatdenov, B G

    2002-01-01

    We consider a process of quasielastic e\\mu large-angle scattering at high energies with radiative corrections up to a two-loop level. The lowest order radiative correction arising both from one-loop virtual photon emission and a real soft emission are presented to a power accuracy. Two-loop level corrections are supposed to be of three gauge-invariant classes. One of them, so-called vertex contribution, is given in logarithmic approximation. Relation with the renormalization group approach is discussed.

  4. (Quasi)Elastic Electron-Muon Large-Angle Scattering to a Two-Loop Approximation Vertex Contributions

    CERN Document Server

    Bytev, V V; Shaikhatdenov, B G

    2002-01-01

    We consider a process of quasielastic e\\mu large-angle scattering at high energies with radiative corrections up to a two-loop level. The lowest order radiative correction arising both from one-loop virtual photon emission and a real soft emission are presented to a power accuracy. Two-loop level corrections are supposed to be of three gauge-invariant classes. One of them, so-called vertex contribution, is given in logarithmic approximation. Relation with the renormalization group approach is discussed.

  5. PWR: 10 years after and perspectives

    International Nuclear Information System (INIS)

    1990-01-01

    These proceedings of the SFEN days on PWR (Ten years after and perspectives) comprise 13 conferences bearing on: - From the occurential approach to the state approach - Evolution of calculating tools - Human factors and safety - Reactor safety in the PWR 2000 - The PWR and the electrical power grid load follow - Fuel aspect of PWR management - PWR chemistry evolution - Balance of radiation protection - PWR modifications balance and influence on reactor operation - Design and maintenance of reactor components: 4 conferences [fr

  6. 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

  7. 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

  8. 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.

  9. 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

  10. 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

  11. 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

  12. 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)

  13. The effects of fission gas release on PWR fuel rod design and performance

    International Nuclear Information System (INIS)

    Leech, W.J.; Kaiser, R.S.

    1980-01-01

    The purpose of this investigation was to determine the effects of fission gas release on PWR fuel rod design and performance. Empirical models were developed from fission gas release data. Fission gas release during normal operation is a function of burnup. There is little additional fission gas release during anticipated transients. The empirical models were used to evaluate Westinghouse fuel rod designs. It was determined that fission gas release is not a limiting parameter for obtaining rod average burnups in the range of 50,000 to 60,000 MWD/MTU. Fission gas release during anticipated transients has a negligible effect on the margins to rod design limits. (author)

  14. Lateral hydraulic forces calculation on PWR fuel assemblies with computational fluid dynamics codes

    International Nuclear Information System (INIS)

    Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.

    2016-01-01

    To be able to simulate the behavior of nuclear fuel under operating conditions, it is required to include all the representative loads, including the lateral hydraulic forces which were not included traditionally because of the difficulty of calculating them in a reliable way. Thanks to the advance in CFD codes, now it is possible to assess them. This study calculates the local lateral hydraulic forces, caused by the contraction and expansion of the flow due to the bow of the surrounding fuel assemblies, on of fuel assembly under typical operating conditions from a three loop Westinghouse PWR reactor. (Author)

  15. VANTAGE 5 PWR fuel assembly demonstration program at Virgil C. Summer nuclear station

    International Nuclear Information System (INIS)

    Warner, D.C.; Orr, W.L.

    1985-01-01

    VANTAGE 5 is an improved PWR fuel product designed and manufactured by Westinghouse Electric Corporation. The VANTAGE 5 fuel design features integral fuel burnable absorbers, intermediate flow mixer grids, axial blankets, high burnup capability, and a reconstitutable top nozzle. A demonstration program for this fuel design commenced in late 1984 in cycle 2 of the Virgil C. Summer Nuclear Station. Objectives for VANTAGE 5 fuel are reduced fuel cycle costs, better core operating margins, and increased design and operating flexibility. Inspections of the VANTAGE 5 demonstration assemblies are planned at each refueling outage

  16. Naturalness made easy: two-loop naturalness bounds on minimal SM extensions

    Energy Technology Data Exchange (ETDEWEB)

    Clarke, Jackson D.; Cox, Peter [ARC Centre of Excellence for Particle Physics at the Terascale,School of Physics, University of Melbourne,Melbourne, 3010 (Australia)

    2017-02-24

    The main result of this paper is a collection of conservative naturalness bounds on minimal extensions of the Standard Model by (vector-like) fermionic or scalar gauge multiplets. Within, we advocate for an intuitive and physical concept of naturalness built upon the renormalisation group equations. In the effective field theory of the Standard Model plus a gauge multiplet with mass M, the low scale Higgs mass parameter is a calculable function of (MS)-bar input parameters defined at some high scale Λ{sub h}>M. If the Higgs mass is very sensitive to these input parameters, then this signifies a naturalness problem. To sensibly capture the sensitivity, it is shown how a sensitivity measure can be rigorously derived as a Bayesian model comparison, which reduces in a relevant limit to a Barbieri-Giudice-like fine-tuning measure. This measure is fully generalisable to any perturbative EFT. The interesting results of our two-loop renormalisation group study are as follows: for Λ{sub h}=Λ{sub Pl} we find “10% fine-tuning” bounds on the masses of various gauge multiplets of M

  17. BOKASUN: a fast and precise numerical program to calculate the Master Integrals of the two-loop sunrise diagrams

    OpenAIRE

    Caffo, Michele; Czyz, Henryk; Gunia, Michal; Remiddi, Ettore

    2008-01-01

    We present the program BOKASUN for fast and precise evaluation of the Master Integrals of the two-loop self-mass sunrise diagram for arbitrary values of the internal masses and the external four-momentum. We use a combination of two methods: a Bernoulli accelerated series expansion and a Runge-Kutta numerical solution of a system of linear differential equations.

  18. The new lattice code Paragon and its qualification for PWR core applications

    International Nuclear Information System (INIS)

    Ouisloumen, M.; Huria, H.C.; Mayhue, L.T.; Smith, R.M.; Kichty, M.J.; Matsumoto, H.; Tahara, Y.

    2003-01-01

    Paragon is a new two-dimensional transport code based on collision probability with interface current method and written entirely in Fortran 90/95. The qualification of Paragon has been completed and the results are very good. This qualification included a number of critical experiments. Comparisons to the Monte Carlo code MCNP for a wide variety of PWR assembly lattice types were also performed. In addition, Paragon-based core simulator models have been compared against PWR plant startup and operational data for a large number of plants. Some results of these calculations and also comparisons against models developed with a licensed Westinghouse lattice code, Phoenix-P, are presented. The qualification described in this paper provided the basis for the qualification of Paragon both as a validated transport code and as the nuclear data source for core simulator codes

  19. Light water reactors development in Japan. (1) Introduction of LWR technology (PWR)

    International Nuclear Information System (INIS)

    Yamada, Ichita; Suzuki, Shigemitsu

    2008-01-01

    Evolutionary progress of the LWR plants in the last half-century was reviewed in series. Introduction of LWR technology (PWR) in Japan was reviewed in this article. Kansai Electric Power imported the Mihama-1 - a 340 MWe PWR built by Westinghouse Corp. It began operating in 1970 to supply power to the World Exposition (EXPO70). There followed a period in which designs was purchased from US vendors and they were constructed with the co-operation of Mitsubishi Heavy Industry, who would then receive a license to build similar plants in Japan and develop the capacity to design and construct PWRs by itself. Progress of designs, fabrications, project management and construction of PWRs were reviewed from technology transfer to its autonomy age. (T. Tanaka)

  20. French PWR safety philosophy

    International Nuclear Information System (INIS)

    Conte, M.

    1986-05-01

    Increasing knowledge and lessons learned from starting and operating experience of French nuclear power plants, completed by the experience learned from the operation of foreign reactors, has contributed to the improvement of French PWR design and safety philosophy. Based on a deterministic approach, the French safety analysis was progressively completed by a probabilistic approach, each of them having possibilities and limits. As a consequence of the global risk objective set in 1977 for nuclear reactors, safety analysis was extended to the evaluation of events more complex than the conventional ones, and later to the evaluation of the feasibility of the offsite emergency plans in case of severe accidents

  1. PWR decontamination feasibility study

    Energy Technology Data Exchange (ETDEWEB)

    Silliman, P.L.

    1978-12-18

    The decontamination work which has been accomplished is reviewed and it is concluded that it is worthwhile to investigate further four methods for decontamination for future demonstration. These are: dilute chemical; single stage strong chemical; redox processes; and redox/chemical in combination. Laboratory work is recommended to define the agents and processes for demonstration and to determine the effect of the solvents on PWR materials. The feasibility of Indian Point 1 for decontamination demonstrations is discussed, and it is shown that the system components of Indian Point 1 are well suited for use in demonstrations.

  2. PWR decontamination feasibility study

    International Nuclear Information System (INIS)

    Silliman, P.L.

    1978-01-01

    The decontamination work which has been accomplished is reviewed and it is concluded that it is worthwhile to investigate further four methods for decontamination for future demonstration. These are: dilute chemical; single stage strong chemical; redox processes; and redox/chemical in combination. Laboratory work is recommended to define the agents and processes for demonstration and to determine the effect of the solvents on PWR materials. The feasibility of Indian Point 1 for decontamination demonstrations is discussed, and it is shown that the system components of Indian Point 1 are well suited for use in demonstrations

  3. PWR core design calculations

    Energy Technology Data Exchange (ETDEWEB)

    Trkov, A; Ravnik, M; Zeleznik, N [Inst. Jozef Stefan, Ljubljana (Slovenia)

    1992-07-01

    Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [Slovenian] Opisali smo programski paket CORD-2, ki se uporablja pri projektnih izracunih sredice pri upravljanju tlacnovodnega reaktorja. Prikazana je uporaba paketa in racunskih postopkov za tipicne probleme, ki nastopajo pri projektiranju sredice. Primerjava glavnih rezultatov z eksperimentalnimi vrednostmi je predstavljena kot del preveritvenega procesa. (author)

  4. 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.)

  5. 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

  6. 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

  7. Reactor analysis support package (RASP). Volume 7. PWR set-point methodology. Final report

    International Nuclear Information System (INIS)

    Temple, S.M.; Robbins, T.R.

    1986-09-01

    This report provides an overview of the basis and methodology requirements for determining Pressurized Water Reactor (PWR) technical specifications related setpoints and focuses on development of the methodology for a reload core. Additionally, the report documents the implementation and typical methods of analysis used by PWR vendors during the 1970's to develop Protection System Trip Limits (or Limiting Safety System Settings) and Limiting Conditions for Operation. The descriptions of the typical setpoint methodologies are provided for Nuclear Steam Supply Systems as designed and supplied by Babcock and Wilcox, Combustion Engineering, and Westinghouse. The description of the methods of analysis includes the discussion of the computer codes used in the setpoint methodology. Next, the report addresses the treatment of calculational and measurement uncertainties based on the extent to which such information was available for each of the three types of PWR. Finally, the major features of the setpoint methodologies are compared, and the principal effects of each particular methodology on plant operation are summarized for each of the three types of PWR

  8. 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.

  9. 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)

  10. Simulator testing of the Westinghouse aware alarm management system

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-09-01

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

  11. Simulator testing of the Westinghouse aware alarm management system

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  12. Safety considerations of PWR's

    International Nuclear Information System (INIS)

    Arnold, W.H. Jr.

    1977-01-01

    The safety of the central station pressurized water reactor is well established and substantiated by its excellent operating record. Operating data from 55 reactors of this type have established a record of safe operating history unparalleled by any modern large scale industry. The 186 plants under construction require a continuing commitment to maintain this outstanding record. The safety of the PWR has been further verified by the recently completed Reactor Safety Study (''Rasmussen'' Report). Not only has this study confirmed the exceptionally low risk associated with PWR operation, it has also introduced a valuable new tool in the decision making process. PWR designs, utilizing the philosophy of defense in depth, provide the bases for evaluating margins of safety. The design of the reactor coolant system, the containment system, emergency core cooling system and other related systems and components provide substantial margins of safety under both normal and postulated accident conditions even considering simultaneous effects of earthquakes and other environmental phenomena. Margins of safety in the assessment of various postulated accident conditions, with emphasis on the postulated loss of reactor coolant accident (LOCA), have been evaluated in depth as exemplified by the comprehensive ECCS rulemaking hearings followed by imposition of very conservative Nuclear Regulatory Commission requirements. When evaluated on an engineering best estimate approach, the significant margins to safety for a LOCA become more apparent. Extensive test programs have also substantiated margins to safety limits. These programs have included both separate effects and systems tests. Component testing has also been performed to substantiate performance levels under adverse combinations of environmental stress. The importance of utilizing past experience and of optimizing the deployment of incremental resources is self evident. Recent safety concerns have included specific areas such

  13. PWR secondary water chemistry guidelines

    International Nuclear Information System (INIS)

    Bell, M.J.; Blomgren, J.C.; Fackelmann, J.M.

    1982-10-01

    Steam generators in pressurized water reactor (PWR) nuclear power plants have experienced tubing degradation by a variety of corrosion-related mechanisms which depend directly on secondary water chemistry. As a result of this experience, the Steam Generator Owners Group and EPRI have sponsored a major program to provide solutions to PWR steam generator problems. This report, PWR Secondary Water Chemistry Guidelines, in addition to presenting justification for water chemistry control parameters, discusses available analytical methods, data management and surveillance, and the management philosophy required to successfully implement the guidelines

  14. Order α'(two-loop) equivalence of the string equations of motion and the σ-model Weyl invariance conditions

    International Nuclear Information System (INIS)

    Metsaev, R.R.; Tseytlin, A.A.

    1987-01-01

    We prove the on-shell equivalence of the order α' terms in the string effective equations (for the graviton, dilaton and the antisymmetric tensor) to the vanishing of the corresponding (two-loop) terms in the Weyl anomaly coefficients for the general bosonic σ-model. We first determine the α' term in the string effective action starting with the known expression for the 3- and 4-point string amplitudes. Then we compute the two-loop β-function in the general σ-model with the antisymmetric tensor coupling. Special emphasis is made on the renormalization scheme dependence of the β-function. Our result disagrees with the previously known one and cannot be manifestly expressed in terms of the generalized curvature for the connection with torsion. We also prove (to the order α' 2 ) that the parallelizable spaces are solutions of the string equations of motion and establish the complete 3-loop expression for the 'central charge' coefficient. (orig.)

  15. Comment on the prediction of two-loop standard chiral perturbation theory for low-energy ππ scattering

    International Nuclear Information System (INIS)

    Girlanda, L.; Moussallam, B.; Stern, J.; Knecht, M.

    1997-03-01

    Four of the six parameters defining the two-loop ππ scattering amplitude have been determined using Roy dispersion relations. Combining this information with the Standard χ PT expressions, the threshold parameters, low-energy phases and the O(p 4 ) constants l 1 r , l 2 r are obtained. The result reproduces the correct D-waves but it is incompatible with existing Standard χ PT analyses of K 14 form factors beyond one loop. (author)

  16. Effects of two-loop contributions in the pseudofermion functional renormalization group method for quantum spin systems

    Science.gov (United States)

    Rück, Marlon; Reuther, Johannes

    2018-04-01

    We implement an extension of the pseudofermion functional renormalization group method for quantum spin systems that takes into account two-loop diagrammatic contributions. An efficient numerical treatment of the additional terms is achieved within a nested graph construction which recombines different one-loop interaction channels. In order to be fully self-consistent with respect to self-energy corrections, we also include certain three-loop terms of Katanin type. We first apply this formalism to the antiferromagnetic J1-J2 Heisenberg model on the square lattice and benchmark our results against the previous one-loop plus Katanin approach. Even though the renormalization group (RG) equations undergo significant modifications when including the two-loop terms, the magnetic phase diagram, comprising Néel ordered and collinear ordered phases separated by a magnetically disordered regime, remains remarkably unchanged. Only the boundary position between the disordered and the collinear phases is found to be moderately affected by two-loop terms. On the other hand, critical RG scales, which we associate with critical temperatures Tc, are reduced by a factor of ˜2 indicating that the two-loop diagrams play a significant role in enforcing the Mermin-Wagner theorem. Improved estimates for critical temperatures are also obtained for the Heisenberg ferromagnet on the three-dimensional simple cubic lattice where errors in Tc are reduced by ˜34 % . These findings have important implications for the quantum phase diagrams calculated within the previous one-loop plus Katanin approach which turn out to be already well converged.

  17. Two-loop O(ααs) corrections to the on-shell fermion propagator in the standard model

    International Nuclear Information System (INIS)

    Eiras, Dolors; Steinhauser, Matthias

    2006-01-01

    In this paper we consider mixed two-loop electroweak corrections to the top quark propagator in the Standard Model. In particular, we compute the on-shell renormalization constant for the mass and wave function, which constitute building blocks for many physical processes. The results are expressed in terms of master integrals. For the latter practical approximations are derived. In the case of the mass renormalization constant we find agreement with the results in the literature

  18. PWR burnable absorber evaluation

    International Nuclear Information System (INIS)

    Cacciapouti, R.J.; Weader, R.J.; Malone, J.P.

    1995-01-01

    The purpose of the study was to evaluate the relative neurotic efficiency and fuel cycle cost benefits of PWR burnable absorbers. Establishment of reference low-leakage equilibrium in-core fuel management plans for 12-, 18- and 24-month cycles. Review of the fuel management impact of the integral fuel burnable absorber (IFBA), erbium and gadolinium. Calculation of the U 3 O 8 , UF 6 , SWU, fuel fabrication, and burnable absorber requirements for the defined fuel management plans. Estimation of fuel cycle costs of each fuel management plan at spot market and long-term market fuel prices. Estimation of the comparative savings of the different burnable absorbers in dollar equivalent per kgU of fabricated fuel. (author)

  19. PWR systems transient analysis

    International Nuclear Information System (INIS)

    Kennedy, M.F.; Peeler, G.B.; Abramson, P.B.

    1985-01-01

    Analysis of transients in pressurized water reactor (PWR) systems involves the assessment of the response of the total plant, including primary and secondary coolant systems, steam piping and turbine (possibly including the complete feedwater train), and various control and safety systems. Transient analysis is performed as part of the plant safety analysis to insure the adequacy of the reactor design and operating procedures and to verify the applicable plant emergency guidelines. Event sequences which must be examined are developed by considering possible failures or maloperations of plant components. These vary in severity (and calculational difficulty) from a series of normal operational transients, such as minor load changes, reactor trips, valve and pump malfunctions, up to the double-ended guillotine rupture of a primary reactor coolant system pipe known as a Large Break Loss of Coolant Accident (LBLOCA). The focus of this paper is the analysis of all those transients and accidents except loss of coolant accidents

  20. PWR degraded core analysis

    International Nuclear Information System (INIS)

    Gittus, J.H.

    1982-04-01

    A review is presented of the various phenomena involved in degraded core accidents and the ensuing transport of fission products from the fuel to the primary circuit and the containment. The dominant accident sequences found in the PWR risk studies published to date are briefly described. Then chapters deal with the following topics: the condition and behaviour of water reactor fuel during normal operation and at the commencement of degraded core accidents; the generation of hydrogen from the Zircaloy-steam and the steel-steam reactions; the way in which the core deforms and finally melts following loss of coolant; debris relocation analysis; containment integrity; fission product behaviour during a degraded core accident. (U.K.)

  1. Steam generators in PWR's

    International Nuclear Information System (INIS)

    Michel, R.

    1974-01-01

    The steam generator of the PWR operates according to the principle of natural circulation. It consists of a U-shaped tube bundle whose free ends are welded to a bottom plate. The tube bundle is surrounded by a cylinder jacket which has slots closely above the bottom or tube plate. The feed water mixed with boiling water enters the tube bundle through these slots. Because of its buoyancy, the steam-water mixture flows upwards. Below the tube plate there are chambers for distributing and collecting pressurized water separated by means of a partition wall. By omitting some tubes, a free alloy is created so that the tubes in the center get sufficient water, too. By asymmetrical arrangement of the partition wall it is further possible to limit the tube alloy only to the inlet side for pressurized water. The flow over the tube plate is thus improved on the inlet side. (DG) [de

  2. Modification of the ANC Nodal Code for analysis of PWR assembly bow

    International Nuclear Information System (INIS)

    Franceschini, Fausto; Fetterman, Robert J.; Little, David C.

    2008-01-01

    Refueling operations at certain PWR cores have revealed fuel assemblies with assembly bow that was higher than expected. As the fuel assemblies bow, the gaps between assemblies change from the uniform nominal configuration. This causes a change in the water volume which affects neutron moderation and thereby power distribution, fuel depletion history, rod internal pressure, etc., with non-trivial impacts on the safety analysis. Westinghouse has developed a new methodology for incorporation of assembly bow in its reload safety analysis package. As part of the new process, the standard Westinghouse reactor physics tool for core analysis, the Advanced Nodal Code ANC, has been modified. The modified ANC, ANCGAP, enables explicit treatment of three-dimensional gap distributions in its neutronic calculations; its accuracy is similar to that of the standard ANC, as demonstrated through an extensive benchmark campaign conducted over a variety of fuel compositions and challenging gap configurations. These features make ANCGAP a crucial tool in the Westinghouse assembly bow package. (authors)

  3. Modification of the ANC Nodal Code for analysis of PWR assembly bow

    Energy Technology Data Exchange (ETDEWEB)

    Franceschini, Fausto; Fetterman, Robert J.; Little, David C. [Westinghouse Electric Company LLC, Pittsburgh PA (United States)

    2008-07-01

    Refueling operations at certain PWR cores have revealed fuel assemblies with assembly bow that was higher than expected. As the fuel assemblies bow, the gaps between assemblies change from the uniform nominal configuration. This causes a change in the water volume which affects neutron moderation and thereby power distribution, fuel depletion history, rod internal pressure, etc., with non-trivial impacts on the safety analysis. Westinghouse has developed a new methodology for incorporation of assembly bow in its reload safety analysis package. As part of the new process, the standard Westinghouse reactor physics tool for core analysis, the Advanced Nodal Code ANC, has been modified. The modified ANC, ANCGAP, enables explicit treatment of three-dimensional gap distributions in its neutronic calculations; its accuracy is similar to that of the standard ANC, as demonstrated through an extensive benchmark campaign conducted over a variety of fuel compositions and challenging gap configurations. These features make ANCGAP a crucial tool in the Westinghouse assembly bow package. (authors)

  4. Perturbative study of the QCD phase diagram for heavy quarks at nonzero chemical potential: Two-loop corrections

    Science.gov (United States)

    Maelger, J.; Reinosa, U.; Serreau, J.

    2018-04-01

    We extend a previous investigation [U. Reinosa et al., Phys. Rev. D 92, 025021 (2015), 10.1103/PhysRevD.92.025021] of the QCD phase diagram with heavy quarks in the context of background field methods by including the two-loop corrections to the background field effective potential. The nonperturbative dynamics in the pure-gauge sector is modeled by a phenomenological gluon mass term in the Landau-DeWitt gauge-fixed action, which results in an improved perturbative expansion. We investigate the phase diagram at nonzero temperature and (real or imaginary) chemical potential. Two-loop corrections yield an improved agreement with lattice data as compared to the leading-order results. We also compare with the results of nonperturbative continuum approaches. We further study the equation of state as well as the thermodynamic stability of the system at two-loop order. Finally, using simple thermodynamic arguments, we show that the behavior of the Polyakov loops as functions of the chemical potential complies with their interpretation in terms of quark and antiquark free energies.

  5. Severe accident management development program for VVER-1000 and VVER-440/213 based on the westinghouse owners group approach

    International Nuclear Information System (INIS)

    Felix, E.; Dessars, N.

    2003-01-01

    The development of the Westinghouse Owners Group Severe Accident Management Guidelines (WOG SAMG) between 1991 and 1994 was initiated in response to the U.S. Nuclear Regulatory Commission (NRC) requirement for addressing the regulatory severe accident concerns. Hence, the WOG SAMG is designed to interface with other existing procedures at the plant and is used in accident sequences that have progressed to the point where these other procedures are not applicable any longer, i.e. following core damage. The primary purpose of the WOG SAMG is to reach a controlled stable state, which can be declared when fission product releases are controlled, challenges to the confinement fission product boundary have been mitigated, and adequate heat removal is provided to the core and the containment. Although the WOG SAMG is a generic severe accident management guidance developed for use by the entirety of the operating Westinghouse PWR plants, provisions have been made in their development to address specific features of individual plants such as confinement type and the feasibility of reactor cavity flooding. Similarly, the generic SAMG does not address unique plant features and equipment, but rather allows for consideration of plant specific features and strategies. This adaptable approach has led to several SAMG development programs for VVER-1000 and VVER-440 type of power plants, under Westinghouse' s lead. The first of these programs carried out to completion was for Temelin NPP - VVER-1000 - in the first quarter of 2003. Other ongoing programs aim at providing a similar work for VVER-440 design, namely Dukovany, Mochovce and Bohunice NPPs. The challenge of adapting the existing generic WOG material to plants other than PWRs mainly arises for VVER-440 because of important differences in confinement design, making it more vulnerable to ex-vessel phenomena such as hydrogen burn. Also, for both eastern designs, cavity flooding strategy requires special consideration and

  6. 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

  7. 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)

  8. Lateral hydraulic forces calculation on PWR fuel assemblies with computational fluid dynamics codes; Calculo de fuerzas laterales hidraulicas en elementos combustibles tipo PWR con codigos de dinamica de fluidos coputacional

    Energy Technology Data Exchange (ETDEWEB)

    Corpa Masa, R.; Jimenez Varas, G.; Moreno Garcia, B.

    2016-08-01

    To be able to simulate the behavior of nuclear fuel under operating conditions, it is required to include all the representative loads, including the lateral hydraulic forces which were not included traditionally because of the difficulty of calculating them in a reliable way. Thanks to the advance in CFD codes, now it is possible to assess them. This study calculates the local lateral hydraulic forces, caused by the contraction and expansion of the flow due to the bow of the surrounding fuel assemblies, on of fuel assembly under typical operating conditions from a three loop Westinghouse PWR reactor. (Author)

  9. 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

  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. 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.

  12. 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

  13. 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

  14. 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

  15. 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

  16. TASS code topical report. V.2 TASS code validation report for the non-LOCA transient analysis of the CE and Westinghouse type plants

    International Nuclear Information System (INIS)

    Sim, Suk K.; Chang, W. P.; Kim, K. D.; Lee, S. J.; Kim, H. C.; Yoon, H. Y.

    1997-02-01

    The development of TASS 1.0 code has been completed and validated its capability in applying for the licensing transient analyses of the CE and Westinghouse type operating reactors as well as the PWR plants under construction in Korea. The validation of the TASS 1.0 code has been achieved through the comparison calculations of the FSAR transients, loss of AC power transient plant data, load rejection and startup test data for the reference plants as well as the BETHSY loop steam generator tube rupture test data. TASS 1.0 calculation agrees well with the best FSAR transient and shows its capability in simulating plant transient analyses. (author). 12 refs., 32 tabs., 132 figs

  17. Zircaloy-4 corrosion in PWR's

    International Nuclear Information System (INIS)

    Fyfitch, S.; Smalley, W.R.; Roberts, E.

    1985-01-01

    Zircaloy-4 waterside corrosion has been studied extensively in the nuclear industry for a number of years. Following the early crud-related corrosion failures in the Saxton test reactor, Westinghouse undertook numerous programs to minimize crud deposition on fuel rods in power reactors through primary coolant chemistry control. Modern plants today are operating with improved coolant chemistry guidelines, and crud deposition levels are very low in proportion to earlier experience. Zircaloy-4 corrosion under a variety of coolant chemistry, heat flux and exposure conditions has been studied extensively. Experience to date, even in relatively high coolant temperature plants, has indicated that -for both fuel cladding and structural components- Zircaloy-4 waterside corrosion performance has been excellent. Recognizing future industry trends, however, which will result in Zircaloy-4 being subjected to ever increasing corrosion duties, Westinghouse will continue accumulating Zircaloy-4 corrosion experience in large power plants. 13 refs.

  18. Two-Loop Effective Theory Analysis of π (K)→eνe[γ] Branching Ratios

    International Nuclear Information System (INIS)

    Cirigliano, Vincenzo; Rosell, Ignasi

    2007-01-01

    We study the ratios R e/μ (P) ≡Γ(P→eν e [γ])/Γ(P→μν μ [γ]) (P=π, K) in Chiral Perturbation Theory to order e 2 p 4 . We complement the two-loop effective theory results with a matching calculation of the counterterm, finding R e/μ (π) =(1.2352±0.0001)x10 -4 and R e/μ (K) =(2.477±0.001)x10 -5

  19. Soft radiation in heavy-particle pair production: All-order colour structure and two-loop anomalous dimension

    International Nuclear Information System (INIS)

    Beneke, M.; Falgari, P.; Schwinn, C.

    2010-01-01

    We consider the total production cross section of heavy coloured particle pairs in hadronic collisions at the production threshold. We construct a basis in colour space that diagonalizes to all orders in perturbation theory the soft function, which appears in a new factorization formula for the combined resummation of soft gluon and Coulomb gluon effects. This extends recent results on the structure of soft anomalous dimensions and allows us to determine an analytic expression for the two-loop soft anomalous dimension at threshold for all production processes of interest.

  20. Two-loop self-energy in the Lamb shift of the ground and excited states of hydrogenlike ions

    Science.gov (United States)

    Yerokhin, V. A.

    2018-05-01

    The two-loop self-energy correction to the Lamb shift of hydrogenlike ions is calculated for the 1 s , 2 s , and 2 p1 /2 states and nuclear charge numbers Z =30 -100 . The calculation is performed to all orders in the nuclear binding strength parameter Z α . As compared to previous calculations of this correction, numerical accuracy is improved by an order of magnitude and the region of the nuclear charges is extended. An analysis of the Z dependence of the obtained results demonstrates their consistency with the known Z α -expansion coefficients.

  1. Schouten identities for Feynman graph amplitudes; The Master Integrals for the two-loop massive sunrise graph

    International Nuclear Information System (INIS)

    Remiddi, Ettore; Tancredi, Lorenzo

    2014-01-01

    A new class of identities for Feynman graph amplitudes, dubbed Schouten identities, valid at fixed integer value of the dimension d is proposed. The identities are then used in the case of the two-loop sunrise graph with arbitrary masses for recovering the second-order differential equation for the scalar amplitude in d=2 dimensions, as well as a chained set of equations for all the coefficients of the expansions in (d−2). The shift from d≈2 to d≈4 dimensions is then discussed

  2. Two-loop top and bottom Yukawa corrections to the Higgs-boson masses in the complex MSSM

    Science.gov (United States)

    Paßehr, Sebastian; Weiglein, Georg

    2018-03-01

    Results for the two-loop corrections to the Higgs-boson masses of the MSSM with complex parameters of O{( α _t^2+α _tα _b+α _b^2) } from the Yukawa sector in the gauge-less limit are presented. The corresponding self-energies and their renormalization have been obtained in the Feynman-diagrammatic approach. The impact of the new contributions on the Higgs spectrum is investigated. Furthermore, a comparison with an existing result in the limit of the MSSM with real parameters is carried out. The new results will be included in the public code FeynHiggs.

  3. A new deformation of W-infinity and applications to the two-loop WZNW and conformal affine Toda models

    International Nuclear Information System (INIS)

    Aratyn, H.; Ferreira, L.A.; Gomes, J.F.; Zimerman, A.H.

    1992-01-01

    We constructed a center less W-infinity type of algebra in terms of a generator of a center less Virasoro algebra and an Abelian spin-1 current. This algebra conventionally emerges in the study of pseudo-differential operators on a circle or alternatively within KP hierarchy with Watanabe's bracket. Construction used here is based on a special deformation of the algebra w ∞ of area preserving diffeomorphisms of a 2-manifold. We show that this deformation technique applies to the two-loop WZNW and conformal affine Toda models, establishing henceforth W ∞ invariance of these models. (author)

  4. Two-loop top and bottom Yukawa corrections to the Higgs-boson masses in the complex MSSM

    International Nuclear Information System (INIS)

    Passehr, Sebastian; Weiglein, Georg

    2017-05-01

    Results for the two-loop corrections to the Higgs-boson masses of the MSSM with complex parameters of O(α 2 t +α t α b +α 2 b ) from the Yukawa sector in the gauge-less limit are presented. The corresponding self-energies and their renormalization have been obtained in the Feynman-diagrammatic approach. The impact of the new contributions on the Higgs spectrum is investigated. Furthermore, a comparison with an existing result in the limit of the MSSM with real parameters is carried out. The new results will be included in the public code FeynHiggs.

  5. Comparison of computational performance of GA and PSO optimization techniques when designing similar systems - Typical PWR core case

    Energy Technology Data Exchange (ETDEWEB)

    Souza Lima, Carlos A. [Instituto de Engenharia Nuclear - Divisao de Reatores/PPGIEN, Rua Helio de Almeida 75, Cidade Universitaria - Ilha do Fundao, P.O. Box: 68550 - Zip Code: 21941-972, Rio de Janeiro (Brazil); Instituto Politecnico, Universidade do Estado do Rio de Janeiro, Pos-Graduacao em Modelagem Computacional, Rua Alberto Rangel - s/n, Vila Nova, Nova Friburgo, Zip Code: 28630-050, Nova Friburgo (Brazil); Lapa, Celso Marcelo F.; Pereira, Claudio Marcio do N.A. [Instituto de Engenharia Nuclear - Divisao de Reatores/PPGIEN, Rua Helio de Almeida 75, Cidade Universitaria - Ilha do Fundao, P.O. Box: 68550 - Zip Code: 21941-972, Rio de Janeiro (Brazil); Instituto Nacional de Ciencia e Tecnologia de Reatores Nucleares Inovadores (INCT) (Brazil); Cunha, Joao J. da [Eletronuclear Eletrobras Termonuclear - Gerencia de Analise de Seguranca Nuclear, Rua da Candelaria, 65, 7 andar. Centro, Zip Code: 20091-906, Rio de Janeiro (Brazil); Alvim, Antonio Carlos M. [Universidade Federal do Rio de Janeiro, COPPE/Nuclear, Cidade Universitaria - Ilha do Fundao s/n, P.O.Box 68509 - Zip Code: 21945-970, Rio de Janeiro (Brazil); Instituto Nacional de Ciencia e Tecnologia de Reatores Nucleares Inovadores (INCT) (Brazil)

    2011-06-15

    Research highlights: > Performance of PSO and GA techniques applied to similar system design. > This work uses ANGRA1 (two loop PWR) core as a prototype. > Results indicate that PSO technique is more adequate than GA to solve this kind of problem. - Abstract: This paper compares the performance of two optimization techniques, particle swarm optimization (PSO) and genetic algorithm (GA) applied to the design a typical reduced scale two loop Pressurized Water Reactor (PWR) core, at full power in single phase forced circulation flow. This comparison aims at analyzing the performance in reaching the global optimum, considering that both heuristics are based on population search methods, that is, methods whose population (candidate solution set) evolve from one generation to the next using a combination of deterministic and probabilistic rules. The simulated PWR, similar to ANGRA 1 power plant, was used as a case example to compare the performance of PSO and GA. Results from simulations indicated that PSO is more adequate to solve this kind of problem.

  6. Comparison of computational performance of GA and PSO optimization techniques when designing similar systems - Typical PWR core case

    International Nuclear Information System (INIS)

    Souza Lima, Carlos A.; Lapa, Celso Marcelo F.; Pereira, Claudio Marcio do N.A.; Cunha, Joao J. da; Alvim, Antonio Carlos M.

    2011-01-01

    Research highlights: → Performance of PSO and GA techniques applied to similar system design. → This work uses ANGRA1 (two loop PWR) core as a prototype. → Results indicate that PSO technique is more adequate than GA to solve this kind of problem. - Abstract: This paper compares the performance of two optimization techniques, particle swarm optimization (PSO) and genetic algorithm (GA) applied to the design a typical reduced scale two loop Pressurized Water Reactor (PWR) core, at full power in single phase forced circulation flow. This comparison aims at analyzing the performance in reaching the global optimum, considering that both heuristics are based on population search methods, that is, methods whose population (candidate solution set) evolve from one generation to the next using a combination of deterministic and probabilistic rules. The simulated PWR, similar to ANGRA 1 power plant, was used as a case example to compare the performance of PSO and GA. Results from simulations indicated that PSO is more adequate to solve this kind of problem.

  7. 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)

  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. 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

  10. 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

  11. 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

  12. 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

  13. A direct comparison of MELCOR 1.8.3 and MAAP4 results for several PWR ampersand BWR accident sequences

    International Nuclear Information System (INIS)

    Leonard, M.T.; Ashbaugh, S.G.; Cole, R.K.; Bergeron, K.D.; Nagashima, K.

    1996-01-01

    This paper presents a comparison of calculations of severe accident progression for several postulated accident sequences for representative Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) nuclear power plants performed with the MELCOR 1.8.3 and the MAAP4 computer codes. The PWR system examined in this study is a 1100 MWe system similar in design to a Westinghouse 3-loop plant with a large dry containment; the BWR is a 1100 MWe system similar in design to General Electric BWR/4 with a Mark I containment. A total of nine accident sequences were studied with both codes. Results of these calculations are compared to identify major differences in the timing of key events in the calculated accident progression or other important aspects of severe accident behavior, and to identify specific sources of the observed differences

  14. Analytic result for the two-loop six-point NMHV amplitude in N=4 super Yang-Mills theory

    CERN Document Server

    Dixon, Lance J.; Henn, Johannes M.

    2012-01-01

    We provide a simple analytic formula for the two-loop six-point ratio function of planar N = 4 super Yang-Mills theory. This result extends the analytic knowledge of multi-loop six-point amplitudes beyond those with maximal helicity violation. We make a natural ansatz for the symbols of the relevant functions appearing in the two-loop amplitude, and impose various consistency conditions, including symmetry, the absence of spurious poles, the correct collinear behaviour, and agreement with the operator product expansion for light-like (super) Wilson loops. This information reduces the ansatz to a small number of relatively simple functions. In order to fix these parameters uniquely, we utilize an explicit representation of the amplitude in terms of loop integrals that can be evaluated analytically in various kinematic limits. The final compact analytic result is expressed in terms of classical polylogarithms, whose arguments are rational functions of the dual conformal cross-ratios, plus precisely two function...

  15. Two-loop controller for maximizing performance of a grid-connected photovoltaic - fuel cell hybrid power plant

    Science.gov (United States)

    Ro, Kyoungsoo

    The study started with the requirement that a photovoltaic (PV) power source should be integrated with other supplementary power sources whether it operates in a stand-alone or grid-connected mode. First, fuel cells for a backup of varying PV power were compared in detail with batteries and were found to have more operational benefits. Next, maximizing performance of a grid-connected PV-fuel cell hybrid system by use of a two-loop controller was discussed. One loop is a neural network controller for maximum power point tracking, which extracts maximum available solar power from PV arrays under varying conditions of insolation, temperature, and system load. A real/reactive power controller (RRPC) is the other loop. The RRPC meets the system's requirement for real and reactive powers by controlling incoming fuel to fuel cell stacks as well as switching control signals to a power conditioning subsystem. The RRPC is able to achieve more versatile control of real/reactive powers than the conventional power sources since the hybrid power plant does not contain any rotating mass. Results of time-domain simulations prove not only effectiveness of the proposed computer models of the two-loop controller, but also their applicability for use in transient stability analysis of the hybrid power plant. Finally, environmental evaluation of the proposed hybrid plant was made in terms of plant's land requirement and lifetime COsb2 emissions, and then compared with that of the conventional fossil-fuel power generating forms.

  16. 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...

  17. PWR station blackout transient simulation in the INER integral system test facility

    International Nuclear Information System (INIS)

    Liu, T.J.; Lee, C.H.; Hong, W.T.; Chang, Y.H.

    2004-01-01

    Station blackout transient (or TMLB' scenario) in a pressurized water reactor (PWR) was simulated using the INER Integral System Test Facility (IIST) which is a 1/400 volumetrically-scaled reduce-height and reduce-pressure (RHRP) simulator of a Westinghouse three-loop PWR. Long-term thermal-hydraulic responses including the secondary boil-off and the subsequent primary saturation, pressurization and core uncovery were simulated based on the assumptions of no offsite and onsite power, feedwater and operator actions. The results indicate that two-phase discharge is the major depletion mode since it covers 81.3% of the total amount of the coolant inventory loss. The primary coolant inventory has experienced significant re-distribution during a station blackout transient. The decided parameter to avoid the core overheating is not the total amount of the coolant inventory remained in the primary core cooling system but only the part of coolant left in the pressure vessel. The sequence of significant events during transient for the IIST were also compared with those of the ROSA-IV large-scale test facility (LSTF), which is a 1/48 volumetrically-scaled full-height and full-pressure (FHFP) simulator of a PWR. The comparison indicates that the sequence and timing of these events during TMLB' transient studied in the RHRP IIST facility are generally consistent with those of the FHFP LSTF. (author)

  18. Two-loop amplitudes and master integrals for the production of a Higgs boson via a massive quark and a scalar-quark loop

    CERN Document Server

    Anastasiou, C; Bucherer, S; Daleo, A; Kunszt, Zoltán; Anastasiou, Charalampos; Beerli, Stefan; Bucherer, Stefan; Daleo, Alejandro; Kunszt, Zoltan

    2007-01-01

    We compute all two-loop master integrals which are required for the evaluation of next-to-leading order QCD corrections in Higgs boson production via gluon fusion. Many two-loop amplitudes for 2 -> 1 processes in the Standard Model and beyond can be expressed in terms of these integrals using automated reduction techniques. These integrals also form a subset of the master integrals for more complicated 2 -> 2 amplitudes with massive propagators in the loops. As a first application, we evaluate the two-loop amplitude for Higgs boson production in gluon fusion via a massive quark. Our result is the first independent check of the calculation of Spira, Djouadi, Graudenz and Zerwas. We also present for the first time the two-loop amplitude for gg -> h via a massive squark.

  19. 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

  20. 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

  1. 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.

  2. 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

  3. 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.

  4. 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

  5. PWR AXIAL BURNUP PROFILE ANALYSIS

    International Nuclear Information System (INIS)

    J.M. Acaglione

    2003-01-01

    The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B andW 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001)

  6. 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

  7. Two-loop planar master integrals for the production of off-shell vector bosons in hadron collisions

    International Nuclear Information System (INIS)

    Henn, Johannes M.; Melnikov, Kirill; Smirnov, Vladimir A.

    2014-01-01

    We describe the calculation of all planar master integrals that are needed for the computation of NNLO QCD corrections to the production of two off-shell vector bosons in hadron collisions. The most complicated representatives of integrals in this class are the two-loop four-point functions where two external lines are on the light-cone and two other external lines have different invariant masses. We compute these and other relevant integrals analytically using differential equations in external kinematic variables and express our results in terms of Goncharov polylogarithms. The case of two equal off-shellnesses, recently considered in ref. http://dx.doi.org/10.1007/JHEP08(2013)070, appears as a particular case of our general solution

  8. Analytic two-loop results for self-energy- and vertex-type diagrams with one non-zero mass

    International Nuclear Information System (INIS)

    Fleischer, J.; Kotikov, A.V.; Veretin, O.L.

    1999-01-01

    For a large class of two-loop self-energy- and vertex-type diagrams with only one non-zero mass (m) and the vertices also with only one non-zero external momentum squared (q 2 ) the first few expansion coefficients are calculated by the large mass expansion. This allows us to 'guess' the general structure of these coefficients and to verify them in terms of certain classes of 'basis elements', which are essentially harmonic sums. Since for this case with only one non-zero mass the large mass expansion and the Taylor series in terms of q 2 are identical, this approach yields analytic expressions of the Taylor coefficients, from which the diagram can be easily evaluated numerically in a large domain of the complex q 2 -plane by well known methods. It is also possible to sum the Taylor series and present the results in terms of polylogarithms

  9. Dominant two-loop electroweak corrections to the hadroproduction of a pseudoscalar Higgs boson and its photonic decay

    International Nuclear Information System (INIS)

    Brod, J.; Kniehl, B.A.

    2008-01-01

    We present the dominant two-loop electroweak corrections to the partial decay widths to gluon jets and prompt photons of the neutral CP-odd Higgs boson A 0 , with mass M A 0 W , in the two-Higgs-doublet model for low to intermediate values of the ratio tan β=v 2 /v 1 of the vacuum expectation values. They apply as they stand to the production cross sections in hadronic and two-photon collisions, at the Tevatron, the LHC, and a future photon collider. The appearance of three γ 5 matrices in closed fermion loops requires special care in the dimensional regularization of ultraviolet divergences. The corrections are negative and amount to several percent, so that they fully compensate or partly screen the enhancement due to QCD corrections. (orig.)

  10. On the Kählerian symmetries of the two-loop action of the effective string theory

    CERN Document Server

    Ozkurt, S S

    2003-01-01

    Sometimes ago, it has been proposed in a paper by N.Kaloper and K.A.Meissner (\\PR {\\bf D56} (1997) 7940) that if one makes local redefinitions of fields, it does not change the equations of motion (in the redefined fields); however, this comment has not generally been accepted, namely, the redefined fields satisfy different equations of motion. For this reason, in this paper, it is proved that the whole action can be written as a square of the zeroth-order field equations. In this way, we show that any solution of the zeroth-order field equations, which has some K\\"{a}hler symmetry, at the same time, is also a solution of the two-loop equations.

  11. Extending two Higgs doublet models for two-loop neutrino mass generation and one-loop neutrinoless double beta decay

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Zhen, E-mail: liu-zhen@sjtu.edu.cn; Gu, Pei-Hong, E-mail: peihong.gu@sjtu.edu.cn

    2017-02-15

    We extend some two Higgs doublet models, where the Yukawa couplings for the charged fermion mass generation only involve one Higgs doublet, by two singlet scalars respectively carrying a singly electric charge and a doubly electric charge. The doublet and singlet scalars together can mediate a two-loop diagram to generate a tiny Majorana mass matrix of the standard model neutrinos. Remarkably, the structure of the neutrino mass matrix is fully determined by the symmetric Yukawa couplings of the doubly charged scalar to the right-handed leptons. Meanwhile, a one-loop induced neutrinoless double beta decay can arrive at a testable level even if the electron neutrino has an extremely small Majorana mass. We also study other experimental constraints and implications including some rare processes and Higgs phenomenology.

  12. Extending two Higgs doublet models for two-loop neutrino mass generation and one-loop neutrinoless double beta decay

    Directory of Open Access Journals (Sweden)

    Zhen Liu

    2017-02-01

    Full Text Available We extend some two Higgs doublet models, where the Yukawa couplings for the charged fermion mass generation only involve one Higgs doublet, by two singlet scalars respectively carrying a singly electric charge and a doubly electric charge. The doublet and singlet scalars together can mediate a two-loop diagram to generate a tiny Majorana mass matrix of the standard model neutrinos. Remarkably, the structure of the neutrino mass matrix is fully determined by the symmetric Yukawa couplings of the doubly charged scalar to the right-handed leptons. Meanwhile, a one-loop induced neutrinoless double beta decay can arrive at a testable level even if the electron neutrino has an extremely small Majorana mass. We also study other experimental constraints and implications including some rare processes and Higgs phenomenology.

  13. Two-loop top and bottom Yukawa corrections to the Higgs-boson masses in the complex MSSM

    Energy Technology Data Exchange (ETDEWEB)

    Passehr, Sebastian; Weiglein, Georg

    2017-05-15

    Results for the two-loop corrections to the Higgs-boson masses of the MSSM with complex parameters of O(α{sup 2}{sub t}+α{sub t}α{sub b}+α{sup 2}{sub b}) from the Yukawa sector in the gauge-less limit are presented. The corresponding self-energies and their renormalization have been obtained in the Feynman-diagrammatic approach. The impact of the new contributions on the Higgs spectrum is investigated. Furthermore, a comparison with an existing result in the limit of the MSSM with real parameters is carried out. The new results will be included in the public code FeynHiggs.

  14. Effect of component aging on PWR control rod drive systems

    International Nuclear Information System (INIS)

    Grove, E.; Gunther, W.; Sullivan, K.

    1992-01-01

    An aging assessment of PWR control rod drive (CRD) systems has been completed as part of the US NRC Nuclear Plant Aging Research (NPAR) Program. The design, construction, maintenance, and operation of the Babcock ampersand Wilcox (B ampersand W), Combustion Engineering (CE), and Westinghouse (W) systems were evaluated to determine the potential for degradation as each system ages. Operating experience data were evaluated to identify the predominant failure modes, causes, and effects. This, coupled with an assessment of the materials of construction and operating environment, demonstrate that each design is subject to degradation, which if left unchecked, could affect its safety function as the plant ages. An industry survey, conducted with the assistance of EPRI and NUMARC, identified current CRD system maintenance and inspection practices. The results of this survey indicate that some plants have performed system modifications, replaced components, or augmented existing preventive maintenance practices in response to system aging. The survey results also supported the operating experience data, which concluded that the timely replacement of degraded components, prior to failure, was not always possible using existing condition monitoring techniques. The recommendations presented in this study also include a discussion of more advanced monitoring techniques, which provide trendable results capable of detecting aging

  15. PENGEMBANGAN MODEL UNTUK SIMULASI KESELAMATAN REAKTOR PWR 1000 MWe GENERASI III+ MENGGUNAKAN PROGRAM KOMPUTER RELAP5

    Directory of Open Access Journals (Sweden)

    Andi Sofrany Ekariansyah

    2015-04-01

    Full Text Available Reaktor daya PWR AP1000 yang didesain oleh Westinghouse adalah reaktor Generasi III+ pertama yang telah menerima persetujuan desain dari U.S. Nuclear Regulatory Commission (NRC. Saat ini utilitas China telah memulai pembangunan beberapa unit AP1000 di dua tapak terpilih untuk rencana operasi pada 2013-2015. AP1000 sebagai desain PWR berdasarkan teknologi teruji dari desain PWR lainnya yang dibuat oleh Westinghouse dengan penguatan pada sistem keselamatan pasif dengan demikian dapat dipertimbangkan untuk dibangun di Indonesia bila mengacu pada persyaratan pada PP 43/2006 mengenai Perijinan Reaktor Nuklir. Namun demikian, desain tersebut perlu diverifikasi oleh Technical Support Organization (TSO independen sebelum dapat dibangun di Indonesia. Verifikasi dapat dilakukan menggunakan paket program RELAP5 dalam bentuk analisis kecelakaan. Selama ini analisis kecelakaan PLTN dilakukan untuk tipe PWR 1000 MWe dari generasi II atau tipe konvensional. Mengingat saat ini referensi yang menggambarkan teknologi AP1000 yang menyertakan teknologi keselamatan pasif sudah tersedia maka dilakukan kegiatan pemodelan yang nantinya dapat digunakan untuk melakukan analisis kecelakaan. Metode pengembangan model mengacu pada pedoman IAEA yang terdiri dari pengumpulan data instalasi, pengembangan engineering data dan penyusunan input deck, verifikasi dan validasi data input. Model yang berhasil dikembangkan secara umum telah mewakili sistem AP1000 secara keseluruhan dan dianggap sebagai model dasar. Model tersebut telah diverifikasi dan divalidasi dengan data desain yang terdapat pada referensi dimana respon parameter termohidraulika menunjukkan perbedaan hasil ± 3% selain untuk parameter penurunan tekanan teras yang lebih rendah 13%. Sebagai model dasar, input deck yang diperoleh dapat dikembangkan lebih lanjut dengan mengintegrasikan pemodelan sistem keselamatan, sistem proteksi, dan sistem kendali yang spesifik AP1000 untuk keperluan simulasi keselamatan yang lebih

  16. 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

  17. 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

  18. PWR reload safety evaluation methodology

    International Nuclear Information System (INIS)

    Doshi, P.K.; Chapin, D.L.; Love, D.S.

    1993-01-01

    The current practice for WWER safety analysis is to prepare the plant Safety Analysis Report (SAR) for initial plant operation. However, the existing safety analysis is typically not evaluated for reload cycles to confirm that all safety limits are met. In addition, there is no systematic reanalysis or reevaluation of the safety analyses after there have been changes made to the plant. The Westinghouse process is discussed which is in contrast to this and in which the SAR conclusions are re-validated through evaluation and/or analysis of each reload cycle. (Z.S.)

  19. Modernization of the Almaraz, AscO & VandellOs non-1E Control systems during the last decade the Spanish PWR nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Fuente Arias, E. de la; Serrano Jimenez, J.; Madroñal Rodriguez, E.

    2016-07-01

    During the last decade the Spanish PWR nuclear power plants designed by Westinghouse have planned and implemented the modernization of the non-1E Control systems. The driving forces behind the modernization of the original Control Systems are the management of the obsolescence of these systems and the implementation of functional improvements in the plants to increase the Control System reliability and availability. Westinghouse Ovation platform has been used in the modernization of the Reactor Control System, Turbine Control System, Plant Computer and Feedwater Heaters Level and MSR s Drains tanks Level control. Modernizations have been spread through the years in such a way that there is not impact on the outages and the different organizations on the customer and estinghouse can have dedicated teams to work in these projects. (Author)

  20. PWR-to-PWR fuel cycle model using dry process

    International Nuclear Information System (INIS)

    Iqbal, M.; Jeong, Chang Joon; Rho, Gyu Hong

    2002-03-01

    PWR-to-PWR fuel cycle model has been developed to recycle the spent fuel using the dry fabrication process. Two types of fuels were considered; first fuel was based on low initial enrichment with low discharge burnup and second one was based on more initial enrichment with high discharge burnup in PWR. For recycling calculations, the HELIOS code was used, in which all of the available fission products were considered. The decay of 10 years was applied for reuse of the spent fuel. Sensitivity analysis for the fresh feed material enrichment has also been carried out. If enrichment of the mixing material is increased the saving of uranium reserves would be decreased. The uranium saving of low burned fuel increased from 4.2% to 7.4% in fifth recycling step for 5 wt% to 19.00wt% mixing material enrichment. While for high burned fuel, there was no uranium saving, which implies that higher uranium enrichment required than 5 wt%. For mixing of 15 wt% enriched fuel, the required mixing is about 21.0% and 37.0% of total fuel volume for low and high burned fuel, respectively. With multiple recycling, reductions in waste for low and high burned fuel became 80% and 60%, for first recycling, respectively. In this way, waste can be reduced more and the cost of the waste disposal reduction can provide the economic balance

  1. San Onofre PWR Data for Code Validation of MOX Fuel Depletion Analyses - Revision 1

    International Nuclear Information System (INIS)

    Hermann, O.W.

    2000-01-01

    The isotopic composition of mixed-oxide fuel (fabricated with both uranium and plutonium isotopes) discharged from reactors is of interest to the Fissile Material Disposition Program. The validation of depletion codes used to predict isotopic compositions of MOX fuel, similar to studies concerning uranium-only fueled reactors, thus, is very important. The EEI-Westinghouse Plutonium Recycle Demonstration Program was conducted to examine the use of MOX fuel in the San Onofre PWR, Unit I, during cycles 2 and 3. The data, usually required as input to depletion codes, either one-dimensional or lattice codes, were taken from various sources and compiled into this report. Where data were either lacking or determined inadequate, the appropriate data were supplied from other references. The scope of the reactor operations and design data, in addition to the isotopic analyses, was considered to be of sufficient quality for depletion code validation

  2. Risk analysis of highly combustible gas storage, supply, and distribution systems in PWR plants

    International Nuclear Information System (INIS)

    Simion, G.P.; VanHorn, R.L.; Smith, C.L.; Bickel, J.H.; Sattison, M.B.; Bulmahn, K.D.

    1993-06-01

    This report presents the evaluation of the potential safety concerns for pressurized water reactors (PWRs) identified in Generic Safety Issue 106, Piping and the Use of Highly Combustible Gases in Vital Areas. A Westinghouse four-loop PWR plant was analyzed for the risk due to the use of combustible gases (predominantly hydrogen) within the plant. The analysis evaluated an actual hydrogen distribution configuration and conducted several sensitivity studies to determine the potential variability among PWRs. The sensitivity studies were based on hydrogen and safety-related equipment configurations observed at other PWRs within the United States. Several options for improving the hydrogen distribution system design were identified and evaluated for their effect on risk and core damage frequency. A cost/benefit analysis was performed to determine whether alternatives considered were justifiable based on the safety improvement and economics of each possible improvement

  3. Representing Operational Knowledge of PWR Plant by Using Multilevel Flow Modelling

    DEFF Research Database (Denmark)

    Zhang, Xinxin; Lind, Morten; Jørgensen, Sten Bay

    2014-01-01

    situation and support operational decisions. This paper will provide a general MFM model of the primary side in a standard Westinghouse Pressurized Water Reactor ( PWR ) system including sub - systems of Reactor Coolant System, Rod Control System, Chemical and Volume Control System, emergency heat removal......The aim of this paper is to explore the capability of representing operational knowledge by using Multilevel Flow Modelling ( MFM ) methodology. The paper demonstrate s how the operational knowledge can be inserted into the MFM models and be used to evaluate the plant state, identify the current...... systems. And the sub - systems’ functions will be decomposed into sub - models according to different operational situations. An operational model will be developed based on the operating procedure by using MFM symbols and this model can be used to implement coordination rules for organize the utilizati...

  4. PWR design for low doses in the United Kingdom: The present and the future

    Energy Technology Data Exchange (ETDEWEB)

    Zodiates, A.M.; Willcock, A. [PWR Project Group, Knutsford, England (United Kingdom)

    1995-03-01

    The Pressurizer Water Reactor (PWR) design chosen for adoption by Nuclear Electric plc was based on the Westinghouse Standard Nuclear Unit Power Plant System (SNUPPS). This design was developed to meet the United Kingdom (UK) requirements and those improvements are embodied in the Sizewell B plant. Nuclear Electric plc is now looking to the design of the future PWRs to be built in the UK. These PWRs will be based as replicas of the Sizewell B design, but attention will be given to reducing operator doses further. This paper details the approach in operator protection improvements incorporated at Sizewall B, presents the estimated annual collective dose, and identifies the approach being adopted to reduce further operator doses in future plants.

  5. Efficacious of estimatives of thermal-hydraulic conditions of the PWR core by measured parameters

    International Nuclear Information System (INIS)

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

    1985-01-01

    Using ALMOD 3W2 and COBRA IIIP computer codes an evaluation of usual methods of estimatives of heat transfer conditions in the PWR core was made, using variables of the monitored processes. It was done a parametric study in conditions of the permanent regim to verify the influence of variables such as, pressure, temperature and power in the value of critical heat flux. Parameters to prevent the DNB phenomenon in KWU power plants and Westinghouse were calculated and implemented in the ALMOD 3W2 program to estimate the DNBR evolution. It was identified a common origin to both methods and comparing with detailed calculations of the COBRA IIIP code, it was settled limitations in the application of parameters. (M.C.K.) [pt

  6. PWR plant construction in Japan

    International Nuclear Information System (INIS)

    Tamura, Toshifumi

    2002-01-01

    The construction methods based on the experiences on the Nuclear Island, which is a critical path in the total construction schedule, have been studied and reconsidered in order to construct by more reliable and economical method. So various improved construction method are being applied and the duration of construction is being reduced continuously. So various improved construction method are being applied and the duration of construction is being reduced continuously. In this paper, the history of construction of twenty-three (23) PWR Plant, the actual construction methods and schedule of Ohi-3/4, to which the many improved methods were applied during their construction, are introduced mainly with the improved points for previously constructed plants. And also the situation of construction method for the next PWR Plant is simply explained

  7. Corrosion of PWR steam generators

    International Nuclear Information System (INIS)

    Garnsey, R.

    1979-01-01

    Some designs of pressurized water reactor (PWR) steam generators have experienced a variety of corrosion problems which include stress corrosion cracking, tube thinning, pitting, fatigue, erosion-corrosion and support plate corrosion resulting in 'denting'. Large international research programmes have been mounted to investigate the phenomena. The operational experience is reviewed and mechanisms which have been proposed to explain the corrosion damage are presented. The implications for design development and for boiler and feedwater control are discussed. (author)

  8. PWR system reliability improvement activities

    International Nuclear Information System (INIS)

    Yoshikawa, Yuichiro

    1985-01-01

    In Japan lacking in energy resources, it is our basic energy policy to accelerate the development program of nuclear power, thereby reducing our dependence. As referred to in the foregoing, every effort has been exerted on our part to improve the PWR system reliability by dint of the so-called 'HOMEMADE' TQC activities, which is our brain-child as a result of applying to the energy industry the quality control philosophy developed in the field of manufacturing industry

  9. BOKASUN: A fast and precise numerical program to calculate the Master Integrals of the two-loop sunrise diagrams

    Science.gov (United States)

    Caffo, Michele; Czyż, Henryk; Gunia, Michał; Remiddi, Ettore

    2009-03-01

    We present the program BOKASUN for fast and precise evaluation of the Master Integrals of the two-loop self-mass sunrise diagram for arbitrary values of the internal masses and the external four-momentum. We use a combination of two methods: a Bernoulli accelerated series expansion and a Runge-Kutta numerical solution of a system of linear differential equations. Program summaryProgram title: BOKASUN Catalogue identifier: AECG_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AECG_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 9404 No. of bytes in distributed program, including test data, etc.: 104 123 Distribution format: tar.gz Programming language: FORTRAN77 Computer: Any computer with a Fortran compiler accepting FORTRAN77 standard. Tested on various PC's with LINUX Operating system: LINUX RAM: 120 kbytes Classification: 4.4 Nature of problem: Any integral arising in the evaluation of the two-loop sunrise Feynman diagram can be expressed in terms of a given set of Master Integrals, which should be calculated numerically. The program provides a fast and precise evaluation method of the Master Integrals for arbitrary (but not vanishing) masses and arbitrary value of the external momentum. Solution method: The integrals depend on three internal masses and the external momentum squared p. The method is a combination of an accelerated expansion in 1/p in its (pretty large!) region of fast convergence and of a Runge-Kutta numerical solution of a system of linear differential equations. Running time: To obtain 4 Master Integrals on PC with 2 GHz processor it takes 3 μs for series expansion with pre-calculated coefficients, 80 μs for series expansion without pre-calculated coefficients, from a few seconds up to a few minutes for Runge-Kutta method (depending

  10. 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

  11. 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

  12. PWR secondary water chemistry guidelines: Revision 3

    International Nuclear Information System (INIS)

    Lurie, S.; Bucci, G.; Johnson, L.; King, M.; Lamanna, L.; Morgan, E.; Bates, J.; Burns, R.; Eaker, R.; Ward, G.; Linnenbom, V.; Millet, P.; Paine, J.P.; Wood, C.J.; Gatten, T.; Meatheany, D.; Seager, J.; Thompson, R.; Brobst, G.; Connor, W.; Lewis, G.; Shirmer, R.; Gillen, J.; Kerns, M.; Jones, V.; Lappegaard, S.; Sawochka, S.; Smith, F.; Spires, D.; Pagan, S.; Gardner, J.; Polidoroff, T.; Lambert, S.; Dahl, B.; Hundley, F.; Miller, B.; Andersson, P.; Briden, D.; Fellers, B.; Harvey, S.; Polchow, J.; Rootham, M.; Fredrichs, T.; Flint, W.

    1993-05-01

    An effective, state-of-the art secondary water chemistry control program is essential to maximize the availability and operating life of major PWR components. Furthermore, the costs related to maintaining secondary water chemistry will likely be less than the repair or replacement of steam generators or large turbine rotors, with resulting outages taken into account. The revised PWR secondary water chemistry guidelines in this report represent the latest field and laboratory data on steam generator corrosion phenomena. This document supersedes Interim PWR Secondary Water Chemistry Recommendations for IGA/SCC Control (EPRI report TR-101230) as well as PWR Secondary Water Chemistry Guidelines--Revision 2 (NP-6239)

  13. 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)

  14. 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

  15. 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

  16. 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)

  17. 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

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

    International Nuclear Information System (INIS)

    Harrison, D.W.

    1991-01-01

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

  19. 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

  20. Two-loop planar master integrals for Higgs →3 partons with full heavy-quark mass dependence

    International Nuclear Information System (INIS)

    Bonciani, Roberto; Duca, Vittorio Del; Frellesvig, Hjalte; Henn, Johannes M.; Moriello, Francesco; Smirnov, Vladimir A.

    2016-01-01

    We present the analytic computation of all the planar master integrals which contribute to the two-loop scattering amplitudes for Higgs→3 partons, with full heavy-quark mass dependence. These are relevant for the NNLO corrections to fully inclusive Higgs production and to the NLO corrections to Higgs production in association with a jet, in the full theory. The computation is performed using the differential equations method. Whenever possible, a basis of master integrals that are pure functions of uniform weight is used. The result is expressed in terms of one-fold integrals of polylogarithms and elementary functions up to transcendental weight four. Two integral sectors are expressed in terms of elliptic integrals. We show that by introducing a one-dimensional parametrization of the integrals the relevant second order differential equation can be readily solved, and the solution can be expressed to all orders of the dimensional regularization parameter in terms of iterated integrals over elliptic kernels. We express the result for the elliptic sectors in terms of two and three-fold iterated integrals, which we find suitable for numerical evaluations. This is the first time that four-point multiscale Feynman integrals have been computed in a fully analytic way in terms of elliptic integrals.

  1. Two-loop master integrals for the mixed EW-QCD virtual corrections to Drell-Yan scattering

    Energy Technology Data Exchange (ETDEWEB)

    Bonciani, Roberto [' ' La Sapienza' ' Univ., Rome (Italy). Dipt. di Fisica; INFN Sezione Roma (Italy); Di Vita, Stefano [Deutsches Elektronen-Synchrotron (DESY), Hamburg (Germany); Mastrolia, Pierpaolo [Max-Planck-Institut fuer Physik, Muenchen (Germany); Padova Univ. (Italy). Dipt. di Fisica e Astronomia; INFN Sezione di Padova (Italy); Schubert, Ulrich [Max-Planck-Institut fuer Physik, Muenchen (Germany)

    2016-04-15

    We present the calculation of the master integrals needed for the two-loop QCD x EW corrections to q+ anti q → l{sup -}+l{sup +} and q+ anti q{sup '} → l{sup -}+ anti ν, for massless external particles. We treat W and Z bosons as degenerate in mass. We identify three types of diagrams, according to the presence of massive internal lines: the no-mass type, the one-mass type, and the two-mass type, where all massive propagators, when occurring, contain the same mass value. We find a basis of 49 master integrals and evaluate them with the method of the differential equations. The Magnus exponential is employed to choose a set of master integrals that obeys a canonical system of differential equations. Boundary conditions are found either by matching the solutions onto simpler integrals in special kinematic configurations, or by requiring the regularity of the solution at pseudo-thresholds. The canonical master integrals are finally given as Taylor series around d=4 space-time dimensions, up to order four, with coefficients given in terms of iterated integrals, respectively up to weight four.

  2. Two-loop planar master integrals for Higgs →3 partons with full heavy-quark mass dependence

    Energy Technology Data Exchange (ETDEWEB)

    Bonciani, Roberto [Dipartimento di Fisica, Sapienza - Università di Roma,Piazzale Aldo Moro 5, 00185, Rome (Italy); INFN Sezione di Roma, Piazzale Aldo Moro 2, 00185, Rome (Italy); Duca, Vittorio Del [ETH Zurich, Institut fur theoretische Physik, Wolfgang-Paulistr. 27, 8093, Zurich (Switzerland); INFN Laboratori Nazionali di Frascati, 00044 Frascati, Roma (Italy); Frellesvig, Hjalte [Institute of Nuclear and Particle Physics, NCSR Demokritos, Agia Paraskevi, 15310 (Greece); Henn, Johannes M. [PRISMA Cluster of Excellence, Johannes Gutenberg University, 55099 Mainz (Germany); Moriello, Francesco [Dipartimento di Fisica, Sapienza - Università di Roma,Piazzale Aldo Moro 5, 00185, Rome (Italy); INFN Sezione di Roma, Piazzale Aldo Moro 2, 00185, Rome (Italy); ETH Zurich, Institut fur theoretische Physik, Wolfgang-Paulistr. 27, 8093, Zurich (Switzerland); Smirnov, Vladimir A. [Skobeltsyn Institute of Nuclear Physics of Moscow State University, 119991 Moscow (Russian Federation)

    2016-12-19

    We present the analytic computation of all the planar master integrals which contribute to the two-loop scattering amplitudes for Higgs→3 partons, with full heavy-quark mass dependence. These are relevant for the NNLO corrections to fully inclusive Higgs production and to the NLO corrections to Higgs production in association with a jet, in the full theory. The computation is performed using the differential equations method. Whenever possible, a basis of master integrals that are pure functions of uniform weight is used. The result is expressed in terms of one-fold integrals of polylogarithms and elementary functions up to transcendental weight four. Two integral sectors are expressed in terms of elliptic integrals. We show that by introducing a one-dimensional parametrization of the integrals the relevant second order differential equation can be readily solved, and the solution can be expressed to all orders of the dimensional regularization parameter in terms of iterated integrals over elliptic kernels. We express the result for the elliptic sectors in terms of two and three-fold iterated integrals, which we find suitable for numerical evaluations. This is the first time that four-point multiscale Feynman integrals have been computed in a fully analytic way in terms of elliptic integrals.

  3. Minimum throttling feedwater control in VVER-1000 and PWR NPPs

    International Nuclear Information System (INIS)

    Symkin, B.E.; Thaulez, F.

    2004-01-01

    This paper presents an approach for the design and implementation of advanced digital control systems that use a minimum-throttling algorithm for the feedwater control. The minimum-throttling algorithm for the feedwater control, i.e. for the control of steam generators level and of the feedwater pumps speed, is applicable for NPPs with variable speed feedwater pumps. It operates in such a way that the feedwater control valve in the most loaded loop is wide open, steam generator level in this loop being controlled by the feedwater pumps speed, while the feedwater control valves in the other loops are slightly throttling under the action of their control system, to accommodate the slight loop imbalances. This has the advantage of minimizing the valve pressure losses hence minimizing the feedwater pumps power consumption and increasing the net MWe. The benefit has been evaluated for specific plants as being roughly 0.7 and 2.4 MW. The minimum throttling mode has the further advantages of lowering the actuator efforts with potential positive impact in actuator life and of minimizing the feedwater pipelines vibrations. The minimum throttling mode of operation has been developed by the Ukrainian company LvivORGRES. It has been applied with great deal of success on several VVER-1000 NPPs, six units of Zaporizhzha in Ukraine plus, with participation of Westinghouse, Kozloduy 5 and 6 in Bulgaria and South Ukraine 1 to 3 in Ukraine. The concept operates with both ON-OFF valves and true control valves. A study, jointly conducted by Westinghouse and LvivORGRES, is ongoing to demonstrate the applicability of the concept to PWRs having variable speed feedwater pumps and having, or installing, digital feedwater control, standalone or as part of a global digital control system. The implementation of the algorithm at VVER-1000 plants provided both safety improvement and direct commercial benefits. The minimum-throttling algorithm will similarly increase the performance of PWRs. The

  4. AP1000. The PWR revisited

    International Nuclear Information System (INIS)

    Gaio, P.

    2006-01-01

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

  5. Analysis of PWR assembly bow

    International Nuclear Information System (INIS)

    Fetterman, Robert J.; Franceschini, Fausto

    2008-01-01

    Excessive out of core assembly bow has been observed during refueling outages of certain PWRs. Assembly bow can take on a rather complex S-shape, and in other cases C-shape bow is prevalent. Concerns have been raised regarding the impact of the observed assembly bow on the in-core power distribution and the safety analyses supporting the plant operations. In response to these concerns, Westinghouse has developed a comprehensive analysis process for determining the effects of assembly bow on core power distributions and plant operating margins. This methodology has been applied to a particular reactor as part of an overall safety reanalysis completed in support of plant modifications. This paper provides a brief description of the methods used and a summary of the pertinent results. (authors)

  6. Analysis of PWR assembly bow

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

    Excessive out of core assembly bow has been observed during refueling outages of certain PWRs. Assembly bow can take on a rather complex S-shape, and in other cases C-shape bow is prevalent. Concerns have been raised regarding the impact of the observed assembly bow on the in-core power distribution and the safety analyses supporting the plant operations. In response to these concerns, Westinghouse has developed a comprehensive analysis process for determining the effects of assembly bow on core power distributions and plant operating margins. This methodology has been applied to a particular reactor as part of an overall safety reanalysis completed in support of plant modifications. This paper provides a brief description of the methods used and a summary of the pertinent results. (authors)

  7. Investigation on the improvement of genetic algorithm for PWR loading pattern search and its benchmark verification

    International Nuclear Information System (INIS)

    Li Qianqian; Jiang Xiaofeng; Zhang Shaohong

    2009-01-01

    In this study, the age technique, the concepts of relativeness degree and worth function are exploited to improve the performance of genetic algorithm (GA) for PWR loading pattern search. Among them, the age technique endows the algorithm be capable of learning from previous search 'experience' and guides it to do a better search in the vicinity ora local optimal; the introduction of the relativeness degree checks the relativeness of two loading patterns before performing crossover between them, which can significantly reduce the possibility of prematurity of the algorithm; while the application of the worth function makes the algorithm be capable of generating new loading patterns based on the statistics of common features of evaluated good loading patterns. Numerical verification against a loading pattern search benchmark problem ora two-loop reactor demonstrates that the adoption of these techniques is able to significantly enhance the efficiency of the genetic algorithm while improves the quality of the final solution as well. (authors)

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

    International Nuclear Information System (INIS)

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

    1990-11-01

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

  9. 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)

  10. 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)

  11. Westinghouse Hanford Company Pollution Prevention Program Implementation Plan

    International Nuclear Information System (INIS)

    Floyd, B.C.

    1994-10-01

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

  12. Westinghouse Advanced Doped Pellet - Characteristics and irradiation behavior

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

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

    1990-07-01

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

  14. Two-loop N=4 super-Yang-Mills effective action and interaction between D3-branes

    International Nuclear Information System (INIS)

    Buchbinder, I.L.; Petrov, A.Yu.; Tseytlin, A.A.

    2002-01-01

    We compute the leading low-energy term in the planar part of the 2-loop contribution to the effective action of N=4 SYM theory in 4 dimensions, assuming that the gauge group SU(N+1) is broken to SU(N)xU(1) by a constant scalar background X. While the leading 1-loop correction is the familiar c 1 F 4 /vertical bar X vertical bar 4 term, the 2-loop expression starts with c 2 F 6 /vertical bar X vertical bar 8 . The 1-loop constant c 1 is known to be equal to the coefficient of the F 4 term in the Born-Infeld action for a probe D3-brane separated by distance vertical bar X vertical bar from a large number N of coincident D3-branes. We show that the same is true also for the 2-loop constant c 2 : it matches the coefficient of the F 6 term in the D3-brane probe action. In the context of the AdS/CFT correspondence, this agreement suggests a non-renormalization of the coefficient of the F 6 term beyond two loops. Thus the result of hep-th/9706072 about the agreement between the v 6 term in the D0-brane supergravity interaction potential and the corresponding 2-loop term in the (1+0)-dimensional reduction of N=4 SYM theory has indeed a direct generalization to 1+3 dimensions, as conjectured earlier in hep-th/9709087. We also discuss the issue of gauge theory-supergravity correspondence for higher order (F 8 , etc.) terms

  15. Conceptual design of simplified PWR

    International Nuclear Information System (INIS)

    Tabata, Hiroaki

    1996-01-01

    The limited availability for location of nuclear power plant in Japan makes plants with higher power ratings more desirable. Having no intention of constructing medium-sized plants as a next generation standard plant, Japanese utilities are interested in applying passive technologies to large ones. So, Japanese utilities have studied large passive plants based on AP600 and SBWR as alternative future LWRs. In a joint effort to develop a new generation nuclear power plant which is more friendly to operator and maintenance personnel and is economically competitive with alternative sources of power generation, JAPC and Japanese Utilities started the study to modify AP600 and SBWR, in order to accommodate the Japanese requirements. During a six year program up to 1994, basic concepts for 1000 MWe class Simplified PWR (SPWR) and Simplified BWR (SBWR) were developed, though there still remain several areas to be improved. These studies have now stepped into the phase of reducing construction cost and searching for maximum power rating that can be attained by reasonably practical technology. These results also suggest that it is hopeful to develop a large 3-loop passive plant (∼1200 MWe). Since Korea mainly deals with PWR, this paper summarizes SPWR study. The SPWR is jointly studied by JAPC, Japanese PWR Utilities, EdF, WH and Mitsubishi Heavy Industry. Using the AP-600 reference design as a basis, we enlarged the plant size to 3-loops and added engineering features to conform with Japanese practice and Utilities' preference. The SPWR program definitively confirmed the feasibility of a passive plant with an NSSS rating about 1000 MWe and 3 loops. (J.P.N.)

  16. Nuclear criticality safety analysis for the traveller PWR fuel shipping package

    International Nuclear Information System (INIS)

    Vescovi, P.J.; Kent, N.A.; Casado, C.A.

    2004-01-01

    The Traveller PWR fresh fuel shipping package represents a radical departure from conventional PWR fuel package designs. Two immediately noticeable features of the Traveller are that it carries a single fuel assembly instead of two as do other package designs, and that it has built-in moderator, which forms part of the flux-trap system. The criticality safety case shows that the Traveller satisfies both U.S. and IAEA licensing requirements, and demonstrates that the package remains acceptably subcritical under normal conditions and hypothetical accident conditions of transport. This paper looks at the modeling techniques that were used to analyze the several accident scenarios that were considered, including: Lattice pitch expansion; Lattice pitch expansion along the fuel assembly length; Preferential flooding (selective flooding of different cavities); Differential flooding (varying water levels inside different cavities); Partial flooding (varying water density); Axial rod displacement; o Sensitivity studies of variable foam densities and boron content in packaging; Analysis for carrying loose rods in a rodbox; The criticality safety case for the Traveller proved to be a successful cooperative effort between ENUSA and Westinghouse

  17. Nuclear criticality safety analysis for the traveller PWR fuel shipping package

    Energy Technology Data Exchange (ETDEWEB)

    Vescovi, P.J.; Kent, N.A.; Casado, C.A. [Westinghouse Electric Co., LLC, Columbia, SC (United States)]|[ENUSA Industrias Avanzadas SA, Madrid (Spain)

    2004-07-01

    The Traveller PWR fresh fuel shipping package represents a radical departure from conventional PWR fuel package designs. Two immediately noticeable features of the Traveller are that it carries a single fuel assembly instead of two as do other package designs, and that it has built-in moderator, which forms part of the flux-trap system. The criticality safety case shows that the Traveller satisfies both U.S. and IAEA licensing requirements, and demonstrates that the package remains acceptably subcritical under normal conditions and hypothetical accident conditions of transport. This paper looks at the modeling techniques that were used to analyze the several accident scenarios that were considered, including: Lattice pitch expansion; Lattice pitch expansion along the fuel assembly length; Preferential flooding (selective flooding of different cavities); Differential flooding (varying water levels inside different cavities); Partial flooding (varying water density); Axial rod displacement; o Sensitivity studies of variable foam densities and boron content in packaging; Analysis for carrying loose rods in a rodbox; The criticality safety case for the Traveller proved to be a successful cooperative effort between ENUSA and Westinghouse.

  18. Modeling of the thermo-mechanical behaviour of the PWR fuel

    International Nuclear Information System (INIS)

    Mailhe, P.

    2014-01-01

    This article reviews the various physical phenomena that take place in an irradiated fuel rod and presents the development of the thermo-mechanical codes able to simulate them. Though technically simple the fuel rod is the place where appear 4 types of process: thermal, gas behaviour, mechanical and corrosion that combine involving 5 elements: the fuel pellet, the fuel clad, the fuel-clad gap, the inside volume and the coolant. For instance the pellet is the place where the following mechanical processes took place: thermal dilatation, elastic deformation, creep deformation, densification, solid swelling, gaseous swelling and cracking. The first industrial code simulating the behaviour of the fuel rod was COCCINEL, it was developed by AREVA teams from the American PAD code that was included in the Westinghouse license. Today the GALILEO code has replaced the COPERNIC code that was developed in the beginning of the 2000 years. GALILEO is a synthesis of the state of the art of the different models used in the codes validated for PWR and BWR. GALILEO has been validated on more than 1500 fuel rods concerning PWR, BWR and specific reactors like Siloe, Osiris, HFR, Halden, Studsvik, BR2/3,...) and also for extended burn-ups. (A.C.)

  19. On the two-loop corrections to the pole mass of the B quark in the gaugeless limit of the MSSM

    International Nuclear Information System (INIS)

    Bednyakov, A.V.; Kazakov, D.I.; )

    2007-01-01

    The result for the two-loop corrections to the pole mass of the b quark in the gaugeless limit of the MSSM is presented. In this limit it is assumed that the contribution from the electroweak gauge interactions is small. The result presented here differs from one obtained earlier, especially in some particular regions of the MSSM parameter space [ru

  20. Two-loop RGE of a general renormalizable Yang-Mills theory in a renormalization scheme with an explicit UV cutoff

    Energy Technology Data Exchange (ETDEWEB)

    Chankowski, Piotr H. [Institute of Theoretical Physics, Faculty of Physics, University of Warsaw,Pasteura 5, 02-093 Warsaw (Poland); Lewandowski, Adrian [Max-Planck-Institut für Gravitationsphysik (Albert-Einstein-Institut),Mühlenberg 1, D-14476 Potsdam (Germany); Institute of Theoretical Physics, Faculty of Physics, University of Warsaw,Pasteura 5, 02-093 Warsaw (Poland); Meissner, Krzysztof A. [Institute of Theoretical Physics, Faculty of Physics, University of Warsaw,Pasteura 5, 02-093 Warsaw (Poland)

    2016-11-18

    We perform a systematic one-loop renormalization of a general renormalizable Yang-Mills theory coupled to scalars and fermions using a regularization scheme with a smooth momentum cutoff Λ (implemented through an exponential damping factor). We construct the necessary finite counterterms restoring the BRST invariance of the effective action by analyzing the relevant Slavnov-Taylor identities. We find the relation between the renormalized parameters in our scheme and in the conventional (MS)-bar scheme which allow us to obtain the explicit two-loop renormalization group equations in our scheme from the known two-loop ones in the (MS)-bar scheme. We calculate in our scheme the divergences of two-loop vacuum graphs in the presence of a constant scalar background field which allow us to rederive the two-loop beta functions for parameters of the scalar potential. We also prove that consistent application of the proposed regularization leads to counterterms which, together with the original action, combine to a bare action expressed in terms of bare parameters. This, together with treating Λ as an intrinsic scale of a hypothetical underlying finite theory of all interactions, offers a possibility of an unconventional solution to the hierarchy problem if no intermediate scales between the electroweak scale and the Planck scale exist.

  1. Two-loop massive fermionic operator matrix elements and intial state QED corrections to e{sup +}e{sup -}{yields}{gamma}{sup *}/Z{sup *}

    Energy Technology Data Exchange (ETDEWEB)

    Bluemlein, J. [Deutsches Elektronen-Synchrotron (DESY), Zeuthen (Germany); Freitas, A. de [Deutsches Elektronen-Synchrotron (DESY), Zeuthen (Germany)]|[Universidad Simon Bolivar, Caracas (Venezuela). Dept. de Fisica; Neerven, W. van [Leiden Univ. (Netherlands). Lorentz Institute

    2008-12-15

    We describe the calculation of the two-loop massive operator matrix elements for massive external fermions. These matrix elements are needed for the calculation of the O({alpha}{sup 2}) initial state radiative corrections to e{sup +}e{sup -} annihilation into a neutral virtual gauge boson, based on the renormalization group technique. (orig.)

  2. Surveillance of vibrations in PWR

    International Nuclear Information System (INIS)

    Espefaelt, R.; Lorenzen, J.; Aakerhielm, F.

    1980-07-01

    The core of a PWR - including fuel elements, internal structure, control rods and core support structure inside the pressure vessel - is subjected to forces which can cause vibrations. One sensitive means to detect and analyse such vibrations is by means of the noise from incore and excore neutron detector signals. In this project noise recordings have been made on two occasions in the Ringhals 2 plant and the obtained data been analysed using the Studsvik Noise Analysis Program System (SNAPS). The results have been intepreted and a detailed description of the vibrational status of the core and pressure vessel internals has been produced. On the basis of the obtained results it is proposed that neutron signal noise analysis should be performed at each PWR plant in the beginning, middle and end of each fuel cycle and an analysis be made using the methods developed in the project. It would also provide a contribution to a higher degree of preparedness for diagnostic tasks in case of unexpected and abnormal events. (author)

  3. 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

  4. 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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  6. 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

  7. 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

  8. 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.

  9. 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

  10. 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.

  11. PWR thermocouple mechanical sealing structure

    International Nuclear Information System (INIS)

    Shen Qiuping; He Youguang

    1991-08-01

    The PWR in-core temperature detection device, which is one of measures to insure reactor safety operation, is to monitor and diagnose reactor thermal power output and in-core power distribution. The temperature detection device system uses thermocouples as measuring elements with stainless steel protecting sleeves. The thermocouple has a limited service time and should be replaced after its service time has reached. A new sealing device for the thermocouples of reactor in-core temperature detection system has been developed to facilitate replacement. The structure is complete tight under high temperature and pressure without any leakage and seepage, and easy to be assembled or disassembled in radioactive environment. The device is designed to make it possible to replace the thermocouple one by one if necessary. This is a new, simple and practical structure

  12. PWR standardization: The French experience

    International Nuclear Information System (INIS)

    Bacher, P.E.

    1987-01-01

    After a short historical review of the French PWR programme with 45000 MWe in operation and 15000 MWe under construction, the paper first develops the objectives and limits of the standardizatoin policy. Implementation of standardization is described through successive reactor series and feedback of experience, together with its impact on safety and on codes and standards. Present benefits of standardization range from low engineering costs to low backfitting costs, via higher quality, reduction in construction times and start-up schedules and improved training of operators. The future of the French programme into the 1990's is again with an advanced standardized series, the N4-1400 MW plant. There is no doubt that the very positive experience with standardization is relevant to any country trying to achieve self-reliance in the nuclear power field. (author)

  13. 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.

  14. 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.

  15. Two loop O(αsGFmt2) corrections to the fermionic decay rates of the standard-model Higgs boson

    International Nuclear Information System (INIS)

    Kniehl, B.A.

    1994-05-01

    Low- and intermediate mass Higgs bosons decay preferably into fermion pairs. The one-loop electroweak corrections to the respective decay rates are dominated by a flavour-independent term of O(G F m t 2 ). We calculate the two-loop gluon correction to this term. It turns out that this correction screens the leading high-m t behaviour of the one-loop result by roughly 10%. We also present the two-loop QCD correction to the contribution induced by a pair of fourth-generation quarks with arbitrary masses. As expected, the inclusion of the QCD correction considerably reduces the renormalization-scheme dependence of the prediction. (orig.)

  16. Babcock and Wilcox advanced PWR development

    International Nuclear Information System (INIS)

    Kulynych, G.E.; Lemon, J.E.

    1986-01-01

    The Babcock and Wilcox 600 MWe PWR design is discussed. Main features of the new B-600 design are improvements in reactor system configuration, glandless coolant pumps, safety features, core design and steam generators

  17. Hydrogen production in a PWR during LOCA

    International Nuclear Information System (INIS)

    Cassette, P.

    1983-12-01

    The purpose of this paper is to provide information on hydrogen generation during LOCA in French 900 MW PWR power plants. The design basis accident is taken into account as well as more severe accidents assuming failure of emergency systems

  18. 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

  19. 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.)

  20. 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

  1. 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

  2. 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...

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

    International Nuclear Information System (INIS)

    Halsall, M.J.

    1977-03-01

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

  4. Westinghouse 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)

  5. 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.

  6. 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.

  7. PWR vessel flaw distribution development

    International Nuclear Information System (INIS)

    Rosinski, S.T.; Kennedy, E.L.; Foulds, J.R.; Kinsman, K.M.

    1990-01-01

    This paper reports on PWR pressure vessels which operate under NRC rules and regulatory guides intended to prevent failure of the vessels. Plants failing to meet the operating criteria specified under these rules and regulations are required to analytically demonstrate fitness for service in order to continue operation. The initial flaw size or distribution of initial vessel flaws is a key input to the required vessel integrity analyses. However, the flaw distribution assumed in the development of the NRC Regulations and recommended for the plant specific analyses is potentially over-conservative. This is because the distribution is based on the limited amount of vessel inspection data available at the time the criteria were being developed and does not take full advantage of the more recent and reliable domestic vessel inspection results. The U.S. Department of Energy is funding an effort through Sandia National Laboratories to investigate the possibility of developing a new flaw distribution based on the increased amount and improved reliability of domestic vessel inspection data. Results of Phase I of the program indicate that state-of-the-art NDE systems' capabilities are sufficient for development of a new flaw distribution that could ultimately provide life extension benefits over the presently required operating practice

  8. Upgrading of PWR plant simulators

    International Nuclear Information System (INIS)

    Wada, Tomonori; Sasaki, Kazunori; Nakaishi, Hirokazu.

    1989-01-01

    For the education and training of operators in electric power plants, simulators have been employed, and it is well known that their effect is great. There are operation training simulators which simulate the dynamic characteristics of plants and all the machinery and equipment that operators handle, and train the procedure of restoration at the time of abnormality in plants, education simulators which can analyze the dynamic characteristics of plants efficiently in a short time, and offer information by visualizing phenomena with three-dimensional display and others so as to be easily understandable, and forecast simulators which do the analysis forecasting plant behavior at the time of abnormality in plants, and investigate the necessity of the guide for operation procedure and the countermeasures at the time of emergency. In this explanation, the upgrading of operation training simulators which have been put already in training is discussed. The constitution of simulator system and the instructor function, the outline of PWR plant simulation models comprising thermal flow model, pump model, leak model and so on, the techniques of increasing simulator speed, and the example of analysis using the NUPAC code are reported. (K.I.)

  9. PWR secondary water chemistry study

    International Nuclear Information System (INIS)

    Pearl, W.L.; Sawochka, S.G.

    1977-02-01

    Several types of corrosion damage are currently chronic problems in PWR recirculating steam generators. One probable cause of damage is a local high concentration of an aggressive chemical even though only trace levels are present in feedwater. A wide variety of trace chemicals can find their way into feedwater, depending on the sources of condenser cooling water and the specific feedwater treatment. In February 1975, Nuclear Water and Waste Technology Corporation (NWT), was contracted to characterize secondary system water chemistry at five operating PWRs. Plants were selected to allow effects of cooling water chemistry and operating history on steam generator corrosion to be evaluated. Calvert Cliffs 1, Prairie Island 1 and 2, Surry 2, and Turkey Point 4 were monitored during the program. Results to date in the following areas are summarized: (1) plant chemistry variations during normal operation, transients, and shutdowns; (2) effects of condenser leakage on steam generator chemistry; (3) corrosion product transport during all phases of operation; (4) analytical prediction of chemistry in local areas from bulk water chemistry measurements; and (5) correlation of corrosion damage to chemistry variation

  10. Comprehensive exergetic and economic comparison of PWR and hybrid fossil fuel-PWR power plants

    International Nuclear Information System (INIS)

    Sayyaadi, Hoseyn; Sabzaligol, Tooraj

    2010-01-01

    A typical 1000 MW Pressurized Water Reactor (PWR) nuclear power plant and two similar hybrid 1000 MW PWR plants operate with natural gas and coal fired fossil fuel superheater-economizers (Hybrid PWR-Fossil fuel plants) are compared exergetically and economically. Comparison is performed based on energetic and economic features of three systems. In order to compare system at their optimum operating point, three workable base case systems including the conventional PWR, and gas and coal fired hybrid PWR-Fossil fuel power plants considered and optimized in exergetic and exergoeconomic optimization scenarios, separately. The thermodynamic modeling of three systems is performed based on energy and exergy analyses, while an economic model is developed according to the exergoeconomic analysis and Total Revenue Requirement (TRR) method. The objective functions based on exergetic and exergoeconomic analyses are developed. The exergetic and exergoeconomic optimizations are performed using the Genetic Algorithm (GA). Energetic and economic features of exergetic and exergoeconomic optimized conventional PWR and gas and coal fired Hybrid PWR-Fossil fuel power plants are compared and discussed comprehensively.

  11. Comparison of the phenomenology of SBO sequences with and without seals LOCA Westinghouse PWRs; Comparacion de la fenomenologia de las secuencias de SBO con y sin LOCA de sellos en reactores PWR Westinghouse

    Energy Technology Data Exchange (ETDEWEB)

    Mena Rosell, L.; Queral, C.; Jimenez Varas, G.

    2013-07-01

    SBO sequences have gained notoriety after the accident at Fukushima. Within this type of sequence the appearance or not of seals of the RCP LOCA determines the evolution of the accident. This work has been applied the methodology of integrated safety analysis (ISA), developed by the CSN, sequences of SBO. The objective is to compare the evolution of SBO sequences in a wide spectrum of conditions and recovery times of AC and DC loss. The simulations have been performed with the SCAIS tool coupled to MAAP. The set of simulations carried out, of the order of 2,000 sequences, clearly show the differences in the evolution of sequences with and without seals crazy. This type of analysis allows you to verify which would be the most appropriate management of sequence depending on the appearance or not of the MADWOMAN of seals.

  12. Verification of human actions in SBO sequences with LOCA stamps in Westinghouse PWRs; Verificacion de las actuaciones humanas en secuencias de SBO con LOCA de sellos en reactores PWR Westinghouse

    Energy Technology Data Exchange (ETDEWEB)

    Queral, C.; Mena Rosell, L.; Jimenez Varas, G.

    2013-07-01

    The Fukushima accident has shown the need for tools and methodologies able to analyze human activities and / or capabilities of portable systems that has given the Spanish plants as a result of the stress tests . In this work we have applied the methodology of integrated safety analysis developed by the CSN , to SBO sequences with LOCA stamp. The aim is to show a methodology for testing the performances of the Emergency Operating Procedures and Guides Severe Accident Management. The simulations were performed with the tool SCAIS coupled to MAAP . The results show that there are human activities that may be beneficial in certain sequences but harmful in others. This type of problem is already known and referred to in the GGAS . However, FSR shows a practical way to check human actions cannot be obtained with other methods.

  13. Parallel GPU implementation of PWR reactor burnup

    International Nuclear Information System (INIS)

    Heimlich, A.; Silva, F.C.; Martinez, A.S.

    2016-01-01

    Highlights: • Three GPU algorithms used to evaluate the burn-up in a PWR reactor. • Exhibit speed improvement exceeding 200 times over the sequential. • The C++ container is expansible to accept new nuclides chains. - Abstract: This paper surveys three methods, implemented for multi-core CPU and graphic processor unit (GPU), to evaluate the fuel burn-up in a pressurized light water nuclear reactor (PWR) using the solutions of a large system of coupled ordinary differential equations. The reactor physics simulation of a PWR reactor spends a long execution time with burnup calculations, so performance improvement using GPU can imply in better core design and thus extended fuel life cycle. The results of this study exhibit speed improvement exceeding 200 times over the sequential solver, within 1% accuracy.

  14. ABB advanced BWR and PWR fuel

    International Nuclear Information System (INIS)

    Junkrans, S.; Helmersson, S.; Andersson, S.

    1999-01-01

    Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both BWR and PWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter, proven to meet the -6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10x10 BWR fuel, where ABB is the only vendor to date with multi batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of BWR and PWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its customers. (orig.)

  15. Preliminary study on direct recycling of spent PWR fuel in PWR system

    International Nuclear Information System (INIS)

    Waris, Abdul; Nuha; Novitriana; Kurniadi, Rizal; Su'ud, Zaki

    2012-01-01

    Preliminary study on direct recycling of PWR spent fuel to support SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario has been conducted. Several spent PWR fuel compositions in loaded PWR fuel has been evaluated to obtain the criticality of reactor. The reactor can achieve it criticality for U-235 enrichment in the loaded fresh fuel is at least 4.0 a% with the minimum fraction of the spent fuel in the core is 15.0 %. The neutron spectra become harder with the escalating of U-235 enrichment in the loaded fresh fuel as well as the amount of the spent fuel in the core.

  16. Improvement of availability of PWR nuclear plants through the reduction of the time required for refueling/maintenance outages

    International Nuclear Information System (INIS)

    Mayers, J.B.; Soth, L.G.

    1978-04-01

    The objective of the project, conducted by Commonwealth Research Corporation and Westinghouse Electric Corporation, is to identify improvements in procedures and equipment which will reduce the time required for refueling/maintenance outages at PWR nuclear power plants. The outage of Commonwealth Edison Zion Station Unit 1 in March through May of 1976 was evaluated to identify those items which caused delays and those work activities that offer the potential for significant improvements that could reduce the overall duration of the outage and achieve an improvement in the plant's availability for power production. Modifications in procedures have been developed and were evaluated during one or more outages in 1977. Conceptual designs have been developed for equipment modifications to the refueling system that could reduce the time required for the refueling portion of the outage. The purpose of the interim report is to describe those conceptual designs and to assess their impact upon future outages. Recommendations are included for the implementation of these equipment improvements in a continuation of this program as a demonstration of plant availability benefits that can be realized in PWR nuclear plants already in operation or under construction

  17. Experimental study on secondary depressurization action for PWR vessel bottom small break LOCA with HPI failure and gas inflow (ROSA-V/LSTF test SB-PV-03)

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

    2005-06-01

    A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which is important in case of high pressure injection (HPI) system failure during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating 4-loop Westinghouse-type PWR (3423 MWt). The experiment, SB-PV-03, simulated a PWR vessel bottom SBLOCA with a rupture of ten instrument-tubes which is equivalent to 0.2% cold leg break. Total HPI failure, non-condensable gas inflow from accumulator injection system (AIS) and operator AM actions on steam generator (SG) secondary depressurization at a rate of -55 K/h and auxiliary feedwater (AFW) supply for 30 minutes were assumed as experiment conditions. It is clarified that the AM actions are effective on primary system depressurization until the end of AIS injection at 1.6 MPa, but thereafter become less effective due to inflow of the non-condensable gas, resulting in delay of low pressure injection (LPI) actuation and whole core heatup under continuous water discharge through the bottom break. The report describes these thermohydraulic phenomena related with transient primary coolant mass and AM actions in addition to estimation of non-condensable gas behavior which affected primary-to-secondary heat transfer. (author)

  18. Probabilistic analysis on the failure of reactivity control for the PWR

    Science.gov (United States)

    Sony Tjahyani, D. T.; Deswandri; Sunaryo, G. R.

    2018-02-01

    The fundamental safety function of the power reactor is to control reactivity, to remove heat from the reactor, and to confine radioactive material. The safety analysis is used to ensure that each parameter is fulfilled during the design and is done by deterministic and probabilistic method. The analysis of reactivity control is important to be done because it will affect the other of fundamental safety functions. The purpose of this research is to determine the failure probability of the reactivity control and its failure contribution on a PWR design. The analysis is carried out by determining intermediate events, which cause the failure of reactivity control. Furthermore, the basic event is determined by deductive method using the fault tree analysis. The AP1000 is used as the object of research. The probability data of component failure or human error, which is used in the analysis, is collected from IAEA, Westinghouse, NRC and other published documents. The results show that there are six intermediate events, which can cause the failure of the reactivity control. These intermediate events are uncontrolled rod bank withdrawal at low power or full power, malfunction of boron dilution, misalignment of control rod withdrawal, malfunction of improper position of fuel assembly and ejection of control rod. The failure probability of reactivity control is 1.49E-03 per year. The causes of failures which are affected by human factor are boron dilution, misalignment of control rod withdrawal and malfunction of improper position for fuel assembly. Based on the assessment, it is concluded that the failure probability of reactivity control on the PWR is still within the IAEA criteria.

  19. A Hold-down Margin Assessment using Statistical Method for the PWR Fuel Assembly

    International Nuclear Information System (INIS)

    Jeon, S. Y.; Park, N. K.; Lee, K. S.; Kim, H. K.

    2007-01-01

    The hold-down springs provide an acceptable hold down force against hydraulic uplift force absorbing the length change of the fuel assembly relative to the space between the upper and lower core plates in PWR. These length changes are mainly due to the thermal expansion, irradiation growth and creep down of the fuel assemblies. There are two kinds of hold-down springs depending on the different design concept of the reactor internals of the PWR in Korea, one is a leaf-type hold down spring for Westinghouse type plants and the other is a coil-type hold-down spring for OPR1000 (Optimized Power Reactor 1000). There are four sets of hold-down springs in each fuel assembly for leaf type hold-down spring and each set of the hold-down springs consists of multiple tapered leaves to form a cantilever leaf spring set. The length, width and thickness of the spring leaves are selected to provide the desired spring constant, deflection range, and hold down force. There are four coil springs in each fuel assembly for coil-type hold-down spring. In this study, the hold-down forces and margins were calculated for the leaf-type and coil-type hold-down springs considering geometrical data of the fuel assembly and its components, length changes of the fuel assembly due to thermal expansion, irradiation growth, creep, and irradiation relaxation. The hold-down spring forces were calculated deterministically and statistically to investigate the benefit of the statistical calculation method in view of hold-down margin. The Monte-Carlo simulation method was used for the statistical hold down force calculation

  20. Study On Safety Analysis Of PWR Reactor Core In Transient And Severe Accident Conditions

    International Nuclear Information System (INIS)

    Le Dai Dien; Hoang Minh Giang; Nguyen Thi Thanh Thuy; Nguyen Thi Tu Oanh; Le Thi Thu; Pham Tuan Nam; Tran Van Trung; Le Van Hong; Vo Thi Huong

    2014-01-01

    The cooperation research project on the Study on Safety Analysis of PWR Reactor Core in Transient and Severe Accident Conditions between Institute for Nuclear Science and Technology (INST), VINATOM and Korean Atomic Energy Research Institute (KAERI), Korea has been setup to strengthen the capability of researches in nuclear safety not only in mastering the methods and computer codes, but also in qualifying of young researchers in the field of nuclear safety analysis. Through the studies on the using of thermal hydraulics computer codes like RELAP5, COBRA, FLUENT and CFX the thermal hydraulics research group has made progress in the research including problems for safety analysis of APR1400 nuclear reactor, PIRT methodologies and sub-channel analysis. The study of severe accidents has been started by using MELCOR in collaboration with KAERI experts and the training on the fundamental phenomena occurred in postulated severe accident. For Vietnam side, VVER-1000 nuclear reactor is also intensively studied. The design of core catcher, reactor containment and severe accident management are the main tasks concerning VVER technology. The research results are presented in the 9 th National Conference on Mechanics, Ha Noi, December 8-9, 2012, the 10 th National Conference on Nuclear Science and Technology, Vung Tau, August 14-15, 2013, as well as published in the journal of Nuclear Science and Technology, Vietnam Nuclear Society and other journals. The skills and experience from using computer codes like RELAP5, MELCOR, ANSYS and COBRA in nuclear safety analysis are improved with the nuclear reactors APR1400, Westinghouse 4 loop PWR and especially the VVER-1000 chosen for the specific studies. During cooperation research project, man power and capability of Nuclear Safety center of INST have been strengthen. Three masters were graduated, 2 researchers are engaging in Ph.D course at Hanoi University of Science and Technology and University of Science and Technology, Korea

  1. Constraints on abelian extensions of the Standard Model from two-loop vacuum stability and U(1){sub B−L}

    Energy Technology Data Exchange (ETDEWEB)

    Corianò, Claudio [STAG Research Centre and Mathematical Sciences,University of Southampton, Southampton SO17 1BJ (United Kingdom); Dipartimento di Matematica e Fisica “Ennio De Giorgi' ,Università del Salento and INFN - Sezione di Lecce,Via Arnesano, 73100 Lecce (Italy); Rose, Luigi Delle; Marzo, Carlo [Dipartimento di Matematica e Fisica “Ennio De Giorgi' ,Università del Salento and INFN - Sezione di Lecce,Via Arnesano, 73100 Lecce (Italy)

    2016-02-19

    We present a renormalization group study of the scalar potential in a minimal U(1){sub B−L} extension of the Standard Model involving one extra heavier Higgs and three heavy right-handed neutrinos with family universal B-L charge assignments. We implement a type-I seesaw for the masses of the light neutrinos of the Standard Model. In particular, compared to a previous study, we perform a two-loop extension of the evolution, showing that two-loop effects are essential for the study of the stability of the scalar potential up to the Planck scale. The analysis includes the contribution of the kinetic mixing between the two abelian gauge groups, which is radiatively generated by the evolution, and the one-loop matching conditions at the electroweak scale. By requiring the stability of the potential up to the Planck mass, significant constraints on the masses of the heavy neutrinos, on the gauge couplings and the mixing in the Higgs sector are identified.

  2. Two-Loop Quark Self-Energy in a New Formalism; 2, Renormalization of the Quark Propagator in the Light-Cone Gauge

    CERN Document Server

    Leibbrandt, George; Leibbrandt, George; Williams, Jimmy D.

    2000-01-01

    The complete two-loop correction to the quark propagator, consisting of the spider, rainbow, gluon bubble and quark bubble diagrams, is evaluated in the noncovariant light-cone gauge (lcg). (The overlapping self-energy diagram had already been computed.) The chief technical tools include the powerful matrix integration technique, the n^*-prescription for the spurious poles of 1/qn, and the detailed analysis of the boundary singularities in five- and six-dimensional parameter space. It is shown that the total divergent contribution to the two-loop correction Sigma_2 contains both covariant and noncovariant components, and is a local function of the external momentum p, even off the mass-shell, as all nonlocal divergent terms cancel exactly. Consequently, both the quark mass and field renormalizations are local. The structure of Sigma_2 implies a quark mass counterterm of the form $\\delta m (lcg) = m\\tilde\\alpha_s C_F(3+\\tilde\\alpha_sW) + {\\rm O} (\\tilde\\alpha_s^3)$, the dimensional regulator epsilon, and on th...

  3. Two-loop quark self-energy in a new formalism; 2, Renormalization of the quark propagator in the light-cone gauge

    CERN Document Server

    Leibbrandt, G

    2000-01-01

    For pt.I see ibid., vol.440, p.537-602, 1995. The complete two-loop correction to the quark propagator, consisting of the spider, rainbow, gluon bubble and quark bubble diagrams, is evaluated in the non-covariant light-cone gauge (LCG), n.A/sup a/(x)=0, n/sup 2/=0. (The overlapping self-energy diagram had already been computed.) The chief technical tools include the powerful matrix integration technique, the n*/sub mu /-prescription for the spurious poles of (q.n)/sup -1/, and the detailed analysis of the boundary singularities in five- and six-dimensional parameter space. It is shown that the total divergent contribution to the two-loop correction Sigma /sub 2/ contains both covariant and non-covariant components, and is a local function of the external momentum p, even off the mass-shell, as all non-local divergent terms cancel exactly. Consequently, both the quark mass and field renormalizations are local. The structure of Sigma /sub 2/ implies a quark mass counterterm of the form delta m(LCG)=m alpha /sub...

  4. 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.

  5. 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)

  6. 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

  7. 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

  8. 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

  9. 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

  10. PWR reactors for BBR nuclear power plants

    International Nuclear Information System (INIS)

    Structure and functioning of the nuclear steam generator system developed by BBR and its components are described. Auxiliary systems, control and load following behaviour and fuel management are discussed and the main data of PWR given. The brochure closes with a perspective of the future of the Muelheim-Kaerlich nuclear power plant. (GL) [de

  11. Thermohydraulic calculations of PWR primary circuits

    International Nuclear Information System (INIS)

    Botelho, D.A.

    1984-01-01

    Some mathematical and numerical models from Retran computer codes aiming to simulate reactor transients, are presented. The equations used for calculating one-dimensional flow are integrated using mathematical methods from Flash code, with steam code to correlate the variables from thermodynamic state. The algorithm obtained was used for calculating a PWR reactor. (E.G.) [pt

  12. Reliability of PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Ribeiro, A.A.T.; Muniz, A.A.

    1978-12-01

    Results of the analysis of factors influencing the reliability of international nuclear power plants of the PWR type are presented. The reliability factor is estimated and the probability of its having lower values than a certain specified value is discussed. (Author) [pt

  13. Coolant monitoring systems for PWR reactors

    International Nuclear Information System (INIS)

    Luzhnov, A.M.; Morozov, V.V.; Tsypin, S.G.

    1987-01-01

    The ways of improving information capacity of existing monitoring systems and the necessity of designing new ones for coolant monitoring are reviewed. A wide research program on development of coolant monitoring systems in PWR reactors is analyzed. The possible applications of in-core and out-of-core detectors for coolant monitoring are demonstrated

  14. Secondary systems of PWR and BWR

    International Nuclear Information System (INIS)

    Schindler, N.

    1981-01-01

    The secondary systems of a nuclear power plant comprises the steam, condensate and feedwater cycle, the steam plant auxiliary or ancillary systems and the cooling water systems. The presentation gives a general review about the main systems which show a high similarity of PWR and BWR plants. (orig./RW)

  15. Simulation model of a PWR power plant

    International Nuclear Information System (INIS)

    Larsen, N.

    1987-03-01

    A simulation model of a hypothetical PWR power plant is described. A large number of disturbances and failures in plant function can be simulated. The model is written as seven modules to the modular simulation system for continuous processes DYSIM and serves also as a user example of this system. The model runs in Fortran 77 on the IBM-PC-AT. (author)

  16. Utilization of thorium in PWR type reactors

    International Nuclear Information System (INIS)

    Correa, F.

    1977-01-01

    Uranium 235 consumption is comparatively evaluated with thorium cycle for a PWR type reactor. Modifications are only made in fuels components. U-235 consumption is pratically unchanged in both cycles. Some good results are promised to the mixed U-238/Th-232 fuel cycle in 1/1 proportion [pt

  17. Improvement of PWR reliability by corrosion prevention

    International Nuclear Information System (INIS)

    Takamatsu, Hiroshi

    1999-01-01

    Since first PWR in Japan started commercial operation in 1970, we have encountered the various modes of corrosion on primary and secondary side components. We have paid much efforts for resolving these corrosion problems, that is, investigating the causes of corrosion and establishing the countermeasures for these corrosion. We summarize these efforts in this article. (author)

  18. An evaluation of tight - pitch PWR cores

    International Nuclear Information System (INIS)

    Correa, F.

    1980-01-01

    The subtask of a project carried out at MIT (Massachusetts Institute of Technology) for DOE (Department of Energy) as part of their NASAP/INFCE - related effects involving the optimization of PWR lattices in the recycle model is summarized. (E.G.) [pt

  19. Steam Generator Chemical Cleaning Application: Korean Experience in PWR NPP

    International Nuclear Information System (INIS)

    Hwang, In-Ho; Varrin-Jr, Robert-D.; Little, Michael-J.; Oh, Yeon-Ok; Choo, Seong-Jib; Park, Jin-Hyeok

    2012-09-01

    Korea Hydro and Nuclear Power (KHNP) performed an EPRI/SGOG chemical cleaning of the secondary side of the steam generators at Ulchin Unit 3 (UCN3) in March 2011 and at Ulchin Unit 4 (UCN4) in September 2011. The steam generator chemical cleaning (SGCC) was performed with venting at the top-of-tube sheet (TTS) and at tube support plates (TSPs) 4, 5, 6, 7, 8, 9, and 10. A primary objective of this SGCC was to address outer diameter stress corrosion cracking (ODSCC), which has been observed at the TTS and TSPs in the UCN3 SGs. The EPRI/SGOG process has been shown to effectively reduce prevailing ODSCC rates at the TTS and TSPs, particularly when applied with periodic venting in this application. This was the first full-length SGCC campaign with venting performed in Korea. Ulchin Unit 3 commenced commercial operation in August 1998 and Ulchin Unit 4 commenced commercial operation in December 1999. UCN3 and UCN4 are a two-loop pressurized water reactor (PWR) of the Korea Standard Nuclear Plant (KSNP) design. The SGs contain high-temperature mill annealed (HTMA) Alloy 600 tubing and are similar in design to the Combustion Engineering CE-80. The KSNP SGs have been susceptible to outer diameter stress corrosion cracking (ODSCC), which is consistent with operating experience for other SGs containing Alloy 600HTMA tubing material. The UCN3/4 SGs have recently begun to experience ODSCC. Hankook Jungsoo Industries Co., Ltd (HaJI) was selected as the cleaning vendor by KHNP. To date, HaJI has completed five Advanced Scale Conditioning Agent (ASCA) cleaning applications and two EPRI/SGOG Steam Generator Chemical Cleaning (SGCC) campaigns for KHNP. The goal of total deposit removal of the applications were successfully achieved and the amounts are 3,579 kg at UCN3 and 3,786 kg at UCN4 which values were estimated before each cleaning by analysing ECT signal and liquid samples from the SGs. The deposits from the SGs were primarily composed of magnetite. There were no chemical

  20. MELCOR 1.8.2 assessment: Surry PWR TMLB' (with a DCH study)

    International Nuclear Information System (INIS)

    Kmetyk, L.N.; Cole, R.K. Jr.; Smith, R.C.; Summers, R.M.; Thompson, S.L.

    1994-02-01

    MELCOR is a fully integrated, engineering-level computer code, being developed at Sandia National Laboratories for the USNRC. This code models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRs. As part of an ongoing assessment program, the MELCOR computer code has been used to analyze a station blackout transient in Surry, a three-loop Westinghouse PWR. Basecase results obtained with MELCOR 1.8.2 are presented, and compared to earlier results for the same transient calculated using MELCOR 1.8.1. The effects of new models added in MELCOR 1.8.2 (in particular, hydrodynamic interfacial momentum exchange, core debris radial relocation and core material eutectics, CORSOR-Booth fission product release, high-pressure melt ejection and direct containment heating) are investigated individually in sensitivity studies. The progress in reducing numeric effects in MELCOR 1.8.2, compared to MELCOR 1.8.1, is evaluated in both machine-dependency and time-step studies; some remaining sources of numeric dependencies (valve cycling, material relocation and hydrogen burn) are identified

  1. Risk-Informed External Hazards Analysis for Seismic and Flooding Phenomena for a Generic PWR

    Energy Technology Data Exchange (ETDEWEB)

    Parisi, Carlo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Prescott, Steve [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ma, Zhegang [Idaho National Lab. (INL), Idaho Falls, ID (United States); Spears, Bob [Idaho National Lab. (INL), Idaho Falls, ID (United States); Szilard, Ronaldo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Coleman, Justin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kosbab, Ben [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-07-26

    This report describes the activities performed during the FY2017 for the US-DOE Light Water Reactor Sustainability Risk-Informed Safety Margin Characterization (LWRS-RISMC), Industry Application #2. The scope of Industry Application #2 is to deliver a risk-informed external hazards safety analysis for a representative nuclear power plant. Following the advancements occurred during the previous FYs (toolkits identification, models development), FY2017 focused on: increasing the level of realism of the analysis; improving the tools and the coupling methodologies. In particular the following objectives were achieved: calculation of buildings pounding and their effects on components seismic fragility; development of a SAPHIRE code PRA models for 3-loops Westinghouse PWR; set-up of a methodology for performing static-dynamic PRA coupling between SAPHIRE and EMRALD codes; coupling RELAP5-3D/RAVEN for performing Best-Estimate Plus Uncertainty analysis and automatic limit surface search; and execute sample calculations for demonstrating the capabilities of the toolkit in performing a risk-informed external hazards safety analyses.

  2. Modular 3-D solid finite element model for fatigue analyses of a PWR coolant system

    International Nuclear Information System (INIS)

    Garrido, Oriol Costa; Cizelj, Leon; Simonovski, Igor

    2012-01-01

    Highlights: ► A 3-D model of a reactor coolant system for fatigue usage assessment. ► The performed simulations are a heat transfer and stress analyses. ► The main results are the expected ranges of fatigue loadings. - Abstract: The extension of operational licenses of second generation pressurized water reactor (PWR) nuclear power plants depends to a large extent on the analyses of fatigue usage of the reactor coolant pressure boundary. The reliable estimation of the fatigue usage requires detailed thermal and stress analyses of the affected components. Analyses, based upon the in-service transient loads should be compared to the loads analyzed at the design stage. The thermal and stress transients can be efficiently analyzed using the finite element method. This requires that a 3-D solid model of a given system is discretized with finite elements (FE). The FE mesh density is crucial for both the accuracy and the cost of the analysis. The main goal of the paper is to propose a set of computational tools which assist a user in a deployment of modular spatial FE model of main components of a typical reactor coolant system, e.g., pipes, pressure vessels and pumps. The modularity ensures that the components can be analyzed individually or in a system. Also, individual components can be meshed with different mesh densities, as required by the specifics of the particular transient studied. For optimal accuracy, all components are meshed with hexahedral elements with quadratic interpolation. The performance of the model is demonstrated with simulations performed with a complete two-loop PWR coolant system (RCS). Heat transfer analysis and stress analysis for a complete loading and unloading cycle of the RCS are performed. The main results include expected ranges of fatigue loading for the pipe lines and coolant pump components under the given conditions.

  3. Long-Term Station Blackout Accident Analyses of a PWR with RELAP5/MOD3.3

    Directory of Open Access Journals (Sweden)

    Andrej Prošek

    2013-01-01

    Full Text Available Stress tests performed in Europe after accident at Fukushima Daiichi also required evaluation of the consequences of loss of safety functions due to station blackout (SBO. Long-term SBO in a pressurized water reactor (PWR leads to severe accident sequences, assuming that existing plant means (systems, equipment, and procedures are used for accident mitigation. Therefore the main objective was to study the accident management strategies for SBO scenarios (with different reactor coolant pumps (RCPs leaks assumed to delay the time before core uncovers and significantly heats up. The most important strategies assumed were primary side depressurization and additional makeup water to reactor coolant system (RCS. For simulations of long term SBO scenarios, including early stages of severe accident sequences, the best estimate RELAP5/MOD3.3 and the verified input model of Krško two-loop PWR were used. The results suggest that for the expected magnitude of RCPs seal leak, the core uncovery during the first seven days could be prevented by using the turbine-driven auxiliary feedwater pump and manually depressurizing the RCS through the secondary side. For larger RCPs seal leaks, in general this is not the case. Nevertheless, the core uncovery can be significantly delayed by increasing RCS depressurization.

  4. 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

  5. Evaluation of two loop-mediated isothermal amplification methods for the detection of Salmonella Enteritidis and Listeria monocytogenes in artificially contaminated ready-to-eat fresh products

    Directory of Open Access Journals (Sweden)

    Angeliki Birmpa

    2015-08-01

    Full Text Available In the present study, the effectiveness of two loop-mediated isothermal amplification (LAMP assays was evaluated. Samples of romaine lettuce, strawberries, cherry tomatoes, green onions and sour berries were inoculated with known dilutions (100-108 CFU/g of produce of S. Enteritidis and L. monocytogenes. With LAMP assay, pathogens can be detected in less than 60 min. The limits of detection of S. Enteritidis and L. monocytogenes depended on the food sample tested and on the presence of enrichment step. After enrichment steps, all food samples were found positive even at low initial pathogen levels. The developed LAMP, assays, are expected to become a valuable, robust, innovative, powerful, cheap and fast monitoring tool, which can be extensively used for routine analysis, and screening of contaminated foods by the food industry and the Public Food Health Authorities.

  6. Two loop virtual corrections to b → (d, s)l{sup +}l{sup -} and c → ul{sup +}l{sup -} for arbitrary momentum transfer

    Energy Technology Data Exchange (ETDEWEB)

    Boer, Stefan de [TU Dortmund, Fakultaet fuer Physik, Dortmund (Germany)

    2017-11-15

    Non-factorizable two loop corrections to heavy to light flavor changing neutral current transitions due to matrix elements of current-current operators are calculated analytically for arbitrary momentum transfer. This extends previous work on b → (d, s)l{sup +}l{sup -} transitions. New results for c → ul{sup +}l{sup -} transitions are presented. Recent work on polylogarithms is used for the master integrals. For b → sl{sup +}l{sup -} transitions, the corrections are most significant for the imaginary parts of the effective Wilson coefficients in the large hadronic recoil range. Analytical results and ready-to-use fitted results for a specific set of parameters are provided. (orig.)

  7. Scaling Regimes in the Model of Passive Scalar Advected by the Turbulent Velocity Field with Finite Correlation Time. Influence of Helicity in Two-Loop Approximation

    CERN Document Server

    Chkhetiani, O G; Jurcisinova, E; Jurcisin, M; Mazzino, A; Repasan, M

    2005-01-01

    The advection of a passive scalar quantity by incompressible helical turbulent flow has been investigated in the framework of an extended Kraichnan model. Statistical fluctuations of the velocity field are assumed to have the Gaussian distribution with zero mean and defined noise with finite-time correlation. Actual calculations have been done up to two-loop approximation in the framework of the field-theoretic renormalization group approach. It turned out that the space parity violation (helicity) of a stochastic environment does not affect anomalous scaling which is the peculiar attribute of a corresponding model without helicity. However, stability of asymptotic regimes, where anomalous scaling takes place, and the effective diffusivity strongly depend on the amount of helicity.

  8. The verification of PWR-fuel code for PWR in-core fuel management

    International Nuclear Information System (INIS)

    Surian Pinem; Tagor M Sembiring; Tukiran

    2015-01-01

    In-core fuel management for PWR is not easy because of the number of fuel assemblies in the core as much as 192 assemblies so many possibilities for placement of the fuel in the core. Configuration of fuel assemblies in the core must be precise and accurate so that the reactor operates safely and economically. It is necessary for verification of PWR-FUEL code that will be used in-core fuel management for PWR. PWR-FUEL code based on neutron transport theory and solved with the approach of multi-dimensional nodal diffusion method many groups and diffusion finite difference method (FDM). The goal is to check whether the program works fine, especially for the design and in-core fuel management for PWR. Verification is done with equilibrium core search model at three conditions that boron free, 1000 ppm boron concentration and critical boron concentration. The result of the average burn up fuel assemblies distribution and power distribution at BOC and EOC showed a consistent trend where the fuel with high power at BOC will produce a high burn up in the EOC. On the core without boron is obtained a high multiplication factor because absence of boron in the core and the effect of fission products on the core around 3.8 %. Reactivity effect at 1000 ppm boron solution of BOC and EOC is 6.44 % and 1.703 % respectively. Distribution neutron flux and power density using NODAL and FDM methods have the same result. The results show that the verification PWR-FUEL code work properly, especially for core design and in-core fuel management for PWR. (author)

  9. 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

  10. SCALE6 Hybrid Deterministic-Stochastic Shielding Methodology for PWR Containment Calculations

    International Nuclear Information System (INIS)

    Matijevic, Mario; Pevec, Dubravko; Trontl, Kresimir

    2014-01-01

    The capabilities and limitations of SCALE6/MAVRIC hybrid deterministic-stochastic shielding methodology (CADIS and FW-CADIS) are demonstrated when applied to a realistic deep penetration Monte Carlo (MC) shielding problem of full-scale PWR containment model. The ultimate goal of such automatic variance reduction (VR) techniques is to achieve acceptable precision for the MC simulation in reasonable time by preparation of phase-space VR parameters via deterministic transport theory methods (discrete ordinates SN) by generating space-energy mesh-based adjoint function distribution. The hybrid methodology generates VR parameters that work in tandem (biased source distribution and importance map) in automated fashion which is paramount step for MC simulation of complex models with fairly uniform mesh tally uncertainties. The aim in this paper was determination of neutron-gamma dose rate distribution (radiation field) over large portions of PWR containment phase-space with uniform MC uncertainties. The sources of ionizing radiation included fission neutrons and gammas (reactor core) and gammas from activated two-loop coolant. Special attention was given to focused adjoint source definition which gave improved MC statistics in selected materials and/or regions of complex model. We investigated benefits and differences of FW-CADIS over CADIS and manual (i.e. analog) MC simulation of particle transport. Computer memory consumption by deterministic part of hybrid methodology represents main obstacle when using meshes with millions of cells together with high SN/PN parameters, so optimization of control and numerical parameters of deterministic module plays important role for computer memory management. We investigated the possibility of using deterministic module (memory intense) with broad group library v7 2 7n19g opposed to fine group library v7 2 00n47g used with MC module to fully take effect of low energy particle transport and secondary gamma emission. Compared with

  11. The effect of zinc addition on PWR corrosion product deposition on zircaloy-4

    International Nuclear Information System (INIS)

    Walters, W.S.; Page, J.D.; Gaffka, A.P.; Kingsbury, A.F.; Foster, J.; Anderson, A.; Wickenden, D.; Henshaw, J.; Zmitko, M.; Masarik, V.; Svarc, V.

    2002-01-01

    During the period 1995 to 2001 a programme of loop irradiation tests have been performed to confirm the effectiveness of zinc additions on PWR circuit chemistry and corrosion. The programme included two loop irradiation experiments, and subsequent PIE; the experiments were a baseline test (no added zinc) and a test with added zinc (10 ppb). This paper addresses the findings regarding corrosion product deposition and activation on irradiated Zircaloy-4 surfaces. The findings are relevant to overall corrosion of the reactor primary circuit, the use of zinc as a corrosion inhibitor, and activation and transport of corrosion products. The irradiation experience provides information on the equilibration of the loop chemistry, with deliberate injection of zinc. The PIE used novel and innovative techniques (described below) to obtain samples of the oxide from the irradiated Zircaloy. The results of the PIE, under normal chemistry and zinc chemistry, indicate the effect of zinc on the deposition and activation of corrosion products on Zircaloy. It was found that corrosion product deposition on Zircaloy is enhanced by the addition of zinc (but corrosion product deposition on other materials was reduced in the presence of zinc). Chemical analysis and radioisotope gamma counting results are presented, to interpret the findings. A computer model has also been used to simulate the corrosion product deposition and activation, to assist in the interpretation of the results. (authors)

  12. The PWR spectral code GELS. Pt. 1

    International Nuclear Information System (INIS)

    Penndorf, K.; Schult, F.; Schulz, G.

    1976-01-01

    The code procedures group constant libraries for the static PWR design of whatever fuel cycle - Uranium, Thorium, or Plutonium. The whole reach of temperatures is covered and the treatment of strong lumped absorbers as control or burnable poison pins is included. The main features are: 1) Good accuracy in spite of not fitting the material data to critical experiments; 2) speed and relatively low computer equipment; 3) restriction to PWR's only. In case of demands for higher accuracy there is a further restriction concerning the library data of the epithermal resonance absorbers: They are strictly valid only for several special lattice geometrics. Three samples are given each representing a typical application of the code. Two of them likewise are demonstrations of recalculated experiments. (orig.) [de

  13. Fuel management optimization for a PWR

    International Nuclear Information System (INIS)

    Dumas, M.; Robeau, D.

    1981-04-01

    This study is aimed to optimize the refueling pattern of a PWR. Two methods are developed, they are based on a linearized form of the optimization problem. The first method determines a feasible solution in two steps; in the first one the original problem is replaced by a relaxed one which is solved by the Method of Approximation Programming. The second step is based on the Branch and Bound method to find the feasible solution closest to the solution obtained in the first step. The second method starts from a given refueling pattern and tries to improve this pattern by the calculation of the effects of 2 by 2, 3 by 3 and 4 by 4 permutations on the objective function. Numerical results are given for a typical PWR refueling using the two methods

  14. RSK-guidelines for PWR reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The RSK guidelines for PWA reactors of April 24, 1974, have been revised and amended in this edition. The RSK presents a summary of safety requirements to be observed in the design, construction, and operation of PWR reactors in the form of guidelines. From January 1979 onwards these guidelines will be the basis of siting and safety considerations for new PWR reactors, and newly built nuclear power plants will have to form these guidelines. They are not binding for existing nuclear power plants under construction or in operation. It will be a matter of individual discussion whether or not the guidelines will be applied in these plants. The main purpose of the guidelines is to facilitate discussion among RSK members and to give early information on necessary safety requirements. If the guidelines are observed by producers and operators, the RSK will make statements on individual projects at short notice. (orig./HP) [de

  15. Optimization of reload core design for PWR

    International Nuclear Information System (INIS)

    Shen Wei; Xie Zhongsheng; Yin Banghua

    1995-01-01

    A direct efficient optimization technique has been effected for automatically optimizing the reload of PWR. The objective functions include: maximization of end-of-cycle (EOC) reactivity and maximization of average discharge burnup. The fuel loading optimization and burnable poison (BP) optimization are separated into two stages by using Haling principle. In the first stage, the optimum fuel reloading pattern without BP is determined by the linear programming method using enrichments as control variable, while in the second stage the optimum BP allocation is determined by the flexible tolerance method using the number of BP rods as control variable. A practical and efficient PWR reloading optimization program based on above theory has been encoded and successfully applied to Qinshan Nuclear Power Plant (QNP) cycle 2 reloading design

  16. 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

  17. PWR fuel behavior: lessons learned from LOFT

    International Nuclear Information System (INIS)

    Russell, M.L.

    1981-01-01

    A summary of the experience with the Loss-of-Fluid Test (LOFT) fuel during loss-of-coolant experiments (LOCEs), operational and overpower transient tests and steady-state operation is presented. LOFT provides unique capabilities for obtaining pressurized water reactor (PWR) fuel behavior information because it features the representative thermal-hydraulic conditions which control fuel behavior during transient conditions and an elaborate measurement system to record the history of the fuel behavior

  18. EDF/CIDEN - ONECTRA: PWR decontamination

    International Nuclear Information System (INIS)

    Fayolle, P.; Orcel, H.; Wertz, L.

    2010-01-01

    In the context of PWR circuit renewal (expected in 2011) and their decontamination, an analysis of data coming from cartography and on site decontamination measurements as well as from premise modelling by means of the PANTHERE radioprotection code, is presented. Several French PWRs have been studied. After a presentation of code principles and operation, the authors discuss the radiological context of a workstation, and give an assessment of the annual dose associated with maintenance operations with or without decontamination

  19. EPRI PWR primary water chemistry guidelines revision

    International Nuclear Information System (INIS)

    McElrath, Joel; Fruzzetti, Keith

    2014-01-01

    EPRI periodically updates the PWR Primary Water Chemistry Guidelines as new information becomes available and as required by NEI 97-06 (Steam Generator Program Guidelines) and NEI 03-08 (Guideline for the Management of Materials Issues). The last revision of the PWR water chemistry guidelines identified an optimum primary water chemistry program based on then-current understanding of research and field information. This new revision provides further details with regard to primary water stress corrosion cracking (PWSCC), fuel integrity, and shutdown dose rates. A committee of industry experts, including utility specialists, nuclear steam supply system (NSSS) and fuel vendor representatives, Institute of Nuclear Power Operations (INPO) representatives, consultants, and EPRI staff collaborated in reviewing the available data on primary water chemistry, reactor water coolant system materials issues, fuel integrity and performance issues, and radiation dose rate issues. From the data, the committee updated the water chemistry guidelines that all PWR nuclear plants should adopt. The committee revised guidance with regard to optimization to reflect industry experience gained since the publication of Revision 6. Among the changes, the technical information regarding the impact of zinc injection on PWSCC initiation and dose rate reduction has been updated to reflect the current level of knowledge within the industry. Similarly, industry experience with elevated lithium concentrations with regard to fuel performance and radiation dose rates has been updated to reflect data collected to date. Recognizing that each nuclear plant owner has a unique set of design, operating, and corporate concerns, the guidelines committee has retained a method for plant-specific optimization. Revision 7 of the Pressurized Water Reactor Primary Water Chemistry Guidelines provides guidance for PWR primary systems of all manufacture and design. The guidelines continue to emphasize plant

  20. Optimum fuel use in PWR reactors

    International Nuclear Information System (INIS)

    Neubauer, W.

    1979-07-01

    An optimization program was developed to calculate minimum-cost refuelling schedules for PWR reactors. Optimization was made over several cycles, without any constraints (equilibrium cycle). In developing the optimization program, special consideration was given to an individual treatment of every fuel element and to a sufficiently accurate calculation of all the data required for safe reactor operation. The results of the optimization program were compared with experimental values obtained at Obrigheim nuclear power plant. (orig.) [de

  1. Chemical and radiochemical specifications - PWR power plants

    International Nuclear Information System (INIS)

    Stutzmann, A.

    1997-01-01

    Published by EDF this document gives the chemical specifications of the PWR (Pressurized Water Reactor) nuclear power plants. Among the chemical parameters, some have to be respected for the safety. These parameters are listed in the STE (Technical Specifications of Exploitation). The values to respect, the analysis frequencies and the time states of possible drops are noticed in this document with the motion STE under the concerned parameter. (A.L.B.)

  2. GAIA: AREVAs New PWR fuel assembly design

    Energy Technology Data Exchange (ETDEWEB)

    Vollmert, N.; Gentet, G.; Louf, P.H.; Mindt, M.; O' Brian, J.; Peucker, J.

    2015-07-01

    GAIA is the label of a new PWR Fuel Assembly design developed by AREVA with the objective to provide its customers an advanced fuel assembly design regarding both robustness and performance. Since 2012 GAIA lead fuel assemblies are under irradiation in a Swedish reactor and since 2015 in a U.S. reactor. Visual inspections and examinations carried out so far during the outages confirmed the intended reliability, robustness and the performance enhancement of the design. (Author)

  3. Shielding design for PWR in France

    International Nuclear Information System (INIS)

    Champion, G.; Charransol; Le Dieu de Ville, A.; Nimal, J.C.; Vergnaud, T.

    1983-05-01

    Shielding calculation scheme used in France for PWR is presented here for 900 MWe and 1300 MWe plants built by EDF the French utility giving electricity. Neutron dose rate at areas accessible by personnel during the reactor operation is calculated and compared with the measurements which were carried out in 900 MWe units up to now. Measurements on the first French 1300 MWe reactor are foreseen at the end of 1983

  4. Organization patterns of PWR power plants

    International Nuclear Information System (INIS)

    Leicman, J.

    1980-01-01

    Organization patterns are shown for the St. Lucia 1, North Anna, Sequoyah, and Beaver Valley nuclear power plants, for a typical PWR power plant in the USA, for the Biblis/RWE-KWU nuclear power plants and for a four-unit nuclear power plant operated by Electricite de France as well as for the Loviisa power plant. Organization patterns are also shown for relatively independent and non-independent nuclear power plants according to IAEA recommendations. (J.P.)

  5. Sensitivity analysis of a PWR pressurizer

    International Nuclear Information System (INIS)

    Bruel, Renata Nunes

    1997-01-01

    A sensitivity analysis relative to the parameters and modelling of the physical process in a PWR pressurizer has been performed. The sensitivity analysis was developed by implementing the key parameters and theoretical model lings which generated a comprehensive matrix of influences of each changes analysed. The major influences that have been observed were the flashing phenomenon and the steam condensation on the spray drops. The present analysis is also applicable to the several theoretical and experimental areas. (author)

  6. T Plant removal of PWR Chiller Subsystem

    International Nuclear Information System (INIS)

    Dana, C.M.

    1994-01-01

    The PWR Pool Chiller System is not longer required for support of the Shippingport Blanket Fuel Assemblies Storage. The Engineering Work Plan will provide the overall coordination of the documentation and physical changes to deactivate the unneeded subsystem. The physical removal of all energy sources for the Chiller equipment will be covered under a one time work plan. The documentation changes will be covered using approved Engineering Change Notices and Procedure Change Authorizations as needed

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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)

  13. 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

  14. Nondestructive examination requirements for PWR vessel internals

    International Nuclear Information System (INIS)

    Spanner, J.

    2015-01-01

    This paper describes the requirements for the nondestructive examination of pressurized water reactor (PWR) vessel internals in accordance with the requirements of the EPRI Material Reliability Program (MRP) inspection standard for PWR internals (MRP-228) and the American Society of Mechanical Engineers Section XI In-service Inspection. The MRP vessel internals examinations have been performed at nuclear plants in the USA since 2009. The objective of the inspection standard is to provide the requirements for the nondestructive examination (NDE) methods implemented to support the inspection and evaluation of the internals. The inspection standard contains requirements specific to the inspection methodologies involved as well as requirements for qualification of the NDE procedures, equipment and personnel used to perform the vessel internals inspections. The qualification requirements for the NDE systems will be summarized. Six PWR plants in the USA have completed inspections of their internals using the Inspection and Evaluation Guideline (MRP-227) and the Inspection Standard (MRP-228). Examination results show few instances of service-induced degradation flaws, as expected. The few instances of degradation have mostly occurred in bolting

  15. Modelling activity transport behavior in PWR plant

    International Nuclear Information System (INIS)

    Henshaw, Jim; McGurk, John; Dickinson, Shirley; Burrows, Robert; Hinds, Kelvin; Hussey, Dennis; Deshon, Jeff; Barrios Figueras, Joan Pau; Maldonado Sanchez, Santiago; Fernandez Lillo, Enrique; Garbett, Keith

    2012-09-01

    The activation and transport of corrosion products around a PWR circuit is a major concern to PWR plant operators as these may give rise to high personnel doses. The understanding of what controls dose rates on ex-core surfaces and shutdown releases has improved over the years but still several questions remain unanswered. For example the relative importance of particle and soluble deposition in the core to activity levels in the plant is not clear. Wide plant to plant and cycle to cycle variations are noted with no apparent explanations why such variations are observed. Over the past few years this group have been developing models to simulate corrosion product transport around a PWR circuit. These models form the basis for the latest version of the BOA code and simulate the movement of Fe and Ni around the primary circuit. Part of this development is to include the activation and subsequent transport of radioactive species around the circuit and this paper describes some initial modelling work in this area. A simple model of activation, release and deposition is described and then applied to explain the plant behaviour at Sizewell B and Vandellos II. This model accounts for activation in the core, soluble and particulate activity movement around the circuit and for activity capture ex-core on both the inner and outer oxides. The model gives a reasonable comparison with plant observations and highlights what controls activity transport in these plants and importantly what factors can be ignored. (authors)

  16. Materials performance in operating PWR steam generators

    International Nuclear Information System (INIS)

    Weeks, J.R.

    1975-01-01

    The Inconel-600 tubing in operating PWR steam generators has developed leaks due to intergranular stress corrosion cracking or a general wastage attack, originating from the secondary side of the tubing. Corrosion has been limited to those areas of the steam generators where limited coolant circulation and high heat flux have caused impurities to concentrate. Wastage or pitting attack has always been associated with local concentration of sodium hydrogen phosphates, whereas stress corrosion has been associated with local concentration of sodium or potassium hydroxides. The only instance of stress corrosion originating from the primary side occurred on cold-worked tubing when hydrogen was not added to getter oxygen, and LiOH was not added to raise the pH of the primary coolant. All PWR manufacturers are now recommending that the phosphate treatment of the secondary coolant be abandoned in favor of an all-volatile treatment. Experience in operating plants has shown, however, that removal of phosphate-rich sludge deposits is difficult, and that further wastage and/or intergranular stress corrosion may develop; the residual sodium phosphates gradually convert by reaction with corrosion product hydroxides to sodium hydroxide, which remains concentrated in the limited flow areas. Improvements in circulation patterns have been achieved by inserting flow baffles in some PWR steam generators. Inservice monitoring by eddy current techniques is useful for detecting corrosion-induced defects in the tubing, but irreproducibility in field examinations can lead to uncertainties interpreting the results. (U.S.)

  17. Implementation in free software of the PWR type university nucleo electric simulator (SU-PWR)

    International Nuclear Information System (INIS)

    Valle H, J.; Hidago H, F.; Morales S, J.B.

    2007-01-01

    Presently work is shown like was carried out the implementation of the University Simulator of Nucleo-electric type PWR (SU-PWR). The implementation of the simulator was carried out in a free software simulation platform, as it is Scilab, what offers big advantages that go from the free use and without cost of the product, until the codes modification so much of the system like of the program with the purpose of to improve it or to adapt it to future routines and/or more advanced graphic interfaces. The SU-PWR shows the general behavior of a PWR nuclear plant (Pressurized Water Reactor) describing the dynamics of the plant from the generation process of thermal energy in the nuclear fuel, going by the process of energy transport toward the coolant of the primary circuit the one which in turn transfers this energy to the vapor generators of the secondary circuit where the vapor is expanded by means of turbines that in turn move the electric generator producing in this way the electricity. The pressurizer that is indispensable for the process is also modeled. Each one of these stages were implemented in scicos that is the Scilab tool specialized in the simulation. The simulation was carried out by means of modules that contain the differential equation that mathematically models each stage or equipment of the PWR plant. The result is a series of modules that based on certain entrances and characteristic of the system they generate exits that in turn are the entrance to other module. Because the SU-PWR is an experimental project in early phase, it is even work and modifications to carry out, for what the models that are presented in this work can vary a little the being integrated to the whole system to simulate, but however they already show clearly the operation and the conformation of the plant. (Author)

  18. Activity transport models for PWR primary circuits; PWR-ydinvoimalaitoksen primaeaeripiirin aktiivisuuskulkeutumismallit

    Energy Technology Data Exchange (ETDEWEB)

    Tanner, V; Rosenberg, R [VTT Chemical Technology, Otaniemi (Finland)

    1995-03-01

    The corrosion products activated in the primary circuit form a major source of occupational radiation dose in the PWR reactors. Transport of corrosion activity is a complex process including chemistry, reactor physics, thermodynamics and hydrodynamics. All the mechanisms involved are not known and there is no comprehensive theory for the process, so experimental test loops and plant data are very important in research efforts. Several activity transport modelling attempts have been made to improve the water chemistry control and to minimise corrosion in PWR`s. In this research report some of these models are reviewed with special emphasis on models designed for Soviet VVER type reactors. (51 refs., 16 figs., 4 tabs.).

  19. Program of monitoring PWR fuel in Spain; Programa de Vigilancia de Combustible pwr en Espana

    Energy Technology Data Exchange (ETDEWEB)

    Martinez Murillo, J. C.; Quecedo, M.; Munoz-Roja, C.

    2015-07-01

    In the year 2000 the PWR utilities: Centrales Nucleares Almaraz-Trillo (CNAT) and Asociacion Nuclear Asco-Vandellos (ANAV), and ENUSA Industrias Avanzadas developed and executed a coordinated strategy named PIC (standing for Coordinated Research Program), for achieving the highest level of fuel reliability. The paper will present the scope and results of this program along the years and will summarize the way the changes are managed to ensure fuel integrity. The excellent performance of the ENUSA manufactured fuel in the PWR Spanish NPPs is the best indicator that the expectations on this program are being met. (Author)

  20. 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