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Sample records for inverted pwr core

  1. Thermal hydraulic design of a hydride-fueled inverted PWR core

    International Nuclear Information System (INIS)

    Malen, J.A.; Todreas, N.E.; Hejzlar, P.; Ferroni, P.; Bergles, A.

    2009-01-01

    An inverted PWR core design utilizing U(45%, w/o)ZrH 1.6 fuel (here referred to as U-ZrH 1.6 ) is proposed and its thermal hydraulic performance is compared to that of a standard rod bundle core design also fueled with U-ZrH 1.6 . The inverted design features circular cooling channels surrounded by prisms of fuel. Hence the relative position of coolant and fuel is inverted with respect to the standard rod bundle design. Inverted core designs with and without twisted tape inserts, used to enhance critical heat flux, were analyzed. It was found that higher power and longer cycle length can be concurrently achieved by the inverted core with twisted tape relative to the optimal standard core, provided that higher core pressure drop can be accommodated. The optimal power of the inverted design with twisted tape is 6869 MW t , which is 135% of the optimally powered standard design (5080 MW t -determined herein). Uncertainties in this design regarding fuel and clad dimensions needed to accommodate mechanical loads and fuel swelling are presented. If mechanical and neutronic feasibility of these designs can be confirmed, these thermal assessments imply significant economic advantages for inverted core designs.

  2. 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.)

  3. 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

  4. 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)

  5. 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.)

  6. Conceptual study of advanced PWR core design

    International Nuclear Information System (INIS)

    Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong.

    1997-09-01

    The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs

  7. Conceptual study of advanced PWR core design

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Chang, Moon Hee; Kim, Keung Ku; Joo, Hyung Kuk; Kim, Young Il; Noh, Jae Man; Hwang, Dae Hyun; Kim, Taek Kyum; Yoo, Yon Jong

    1997-09-01

    The purpose of this project is for developing and verifying the core design concepts with enhanced safety and economy, and associated methodologies for core analyses. From the study of the sate-of-art of foreign advanced reactor cores, we developed core concepts such as soluble boron free, high convertible and enhanced safety core loaded semi-tight lattice hexagonal fuel assemblies. To analyze this hexagonal core, we have developed and verified some neutronic and T/H analysis methodologies. HELIOS code was adopted as the assembly code and HEXFEM code was developed for hexagonal core analysis. Based on experimental data in hexagonal lattices and the COBRA-IV-I code, we developed a thermal-hydraulic analysis code for hexagonal lattices. Using the core analysis code systems developed in this project, we designed a 600 MWe core and studied the feasibility of the core concepts. Two additional scopes were performed in this project : study on the operational strategies of soluble boron free core and conceptual design of large scale passive core. By using the axial BP zoning concept and suitable design of control rods, this project showed that it was possible to design a soluble boron free core in 600 MWe PWR. The results of large scale core design showed that passive concepts and daily load follow operation could be practiced. (author). 15 refs., 52 tabs., 101 figs.

  8. Thermal-hydraulic analysis for wire-wrapped PWR cores

    Energy Technology Data Exchange (ETDEWEB)

    Diller, P. [General Electric Company, 3901 Castle Hayne Rd., Wilmington, NC 28401 (United States)], E-mail: pdiller@gmail.com; Todreas, N. [Massachusetts Institute of Technology, Cambridge, MA 02139 (United States)], E-mail: todreas@mit.edu; Hejzlar, P. [Massachusetts Institute of Technology, Cambridge, MA 02139 (United States)

    2009-08-15

    This work focuses on the steady-state and transient thermal-hydraulic analyses for PWR cores using wire wraps in a hexagonal array with either U (45% w/o)-ZrH{sub 1.6} (referred to as U-ZrH{sub 1.6}) or UO{sub 2} fuels. Equivalences (thermal-hydraulic and neutronic) were created between grid spacer and wire wrap designs, and were used to apply results calculated for grid spacers to wire wrap designs. Design limits were placed on the pressure drop, critical heat flux (CHF), fuel and cladding temperature and vibrations. The vibrations limits were imposed for flow-induced vibrations (FIV) and thermal-hydraulic vibrations (THV). The transient analysis examined an overpower accident, loss of coolant accident (LOCA) and loss of flow accident (LOFA). The thermal-hydraulic performance of U-ZrH{sub 1.6} and UO{sub 2} were found very similar. Relative to grid spacer designs, wire wrap designs were found to have smaller fretting wear, substantially lower pressure drop and higher CHF. As a result, wire wrap cores were found to offer substantially higher maximum powers than grid spacer cores, allowing for a 25% power increase relative to the grid spacer uprate [Shuffler, C.A., Malen, J.A., Trant, J.M., Todreas, N.E., 2009a. Thermal-hydraulic analysis for grid supported and inverted fueled PWR cores. Nuclear Technology (this special issue devoted to hydride fuel in LWRs)] and a 58% power increase relative to the reference core.

  9. Evaluation of tight-pitch PWR cores

    International Nuclear Information System (INIS)

    Correa, F.; Driscoll, M.J.; Lanning, D.D.

    1979-08-01

    The impact of tight pinch cores on the consumption of natural uranium ore has been evaluated for two systems of coupled PWR's namely one particular type of thorium system - 235 U/UO 2 : Pu/ThO 2 : 233 U/ThO 2 - and the conventional recycle-mode uranium system - 235 U/UO 2 : Pu/UO 2 . The basic parameter varied was the fuel-to-moderator volume ratio (F/M) of the (uniform) lattice for the last core in each sequence. Although methods and data verification in the range of present interest, 0.5 (current lattices) 1.0, the EPRI-LEOPARD and LASER programs used for the thorium and uranium calculations, respectively, were successfully benchmarked against several of the more pertinent experiments

  10. Economic optimization of PWR cores with ROSA

    International Nuclear Information System (INIS)

    Verhagen, F.C.M.; Wakker, P.H.

    2005-01-01

    The core-loading pattern is decisive for fuel cycle economics, fuel safety parameters and economic planning for future cycles. ROSA, NRG's loading pattern optimization code system for PWRs, has proven for over a decade to be a valuable tool to reactor operators for improving their fuel management economics. ROSA uses simulated annealing as loading pattern optimization technique, in combination with an extremely fast 3-D neutronics code for loading pattern calculations. The code is continuously extended with new optimization parameters and rules. This paper outlines recent developments of the ROSA code system and discusses results of PWR specific applications of ROSA. Core designs with a large variety of challenging constraints have been realized with ROSA. As a typical example, for the 193 assembly, Vantage 5H/RFA-2 fueled TVA's Watts Bar unit 1, a cycle 4 core with 76 feed assemblies was designed. This was followed by a high-energy cycle 5 with only 77 feed assemblies and approximately 535 days of natural cycle length. Subsequently, an economical core using 72 bundles was designed for cycle 6. This resulted in considerable savings in the cost of feed assemblies for reloads. The typical accuracy of ROSA compared to results of license codes in within ±0.02 for normalized assembly powers, ±0.03 for maximum enthalpy rise hot channel factor (F ΔH ), and ±3 days for natural cycle length. (author)

  11. 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)

  12. Cylindrization of a PWR core for neutronic calculations

    International Nuclear Information System (INIS)

    Santos, Rubens Souza dos

    2005-01-01

    In this work we propose a core cylindrization, starting from a PWR core configuration, through the use of an algorithm that becomes the process automated in the program, independent of the discretization. This approach overcomes the problem stemmed from the use of the neutron transport theory on the core boundary, in addition with the singularities associated with the presence of corners on the outer fuel element core of, existents in the light water reactors (LWR). The algorithm was implemented in a computational program used to identification of the control rod drop accident in a typical PWR core. The results showed that the algorithm presented consistent results comparing with an production code, for a problem with uniform properties. In our conclusions, we suggest, for future works, for analyzing the effect on mesh sizes for the Cylindrical geometry, and to compare the transport theory calculations versus diffusion theory, for the boundary conditions with corners, for typical PWR cores. (author)

  13. Two optimal control methods for PWR core control

    International Nuclear Information System (INIS)

    Karppinen, J.; Blomsnes, B.; Versluis, R.M.

    1976-01-01

    The Multistage Mathematical Programming (MMP) and State Variable Feedback (SVF) methods for PWR core control are presented in this paper. The MMP method is primarily intended for optimization of the core behaviour with respect to xenon induced power distribution effects in load cycle operation. The SVF method is most suited for xenon oscillation damping in situations where the core load is unpredictable or expected to stay constant. Results from simulation studies in which the two methods have been applied for control of simple PWR core models are presented. (orig./RW) [de

  14. Characterization of Factors affecting IASCC of PWR Core Internals

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Woo; Hwang, Seong Sik; Kim, Won Sam [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-09-15

    A lot works have been performed on IASCC in BWR. Recent efforts have been devoted to investigate IASCC in PWR, but the mechanism in PWR is not fully understood yet as compared with that in BWR due to a lack of data from laboratories and fields. Therefore it is strongly needed to review and analyse recent researches of IASCC in both BWR and PWR for establishing a proactive management technology for IASCC of core internals in Korean PWRs. This work is aimed to review mainly recent technical reports on IASCC of stainless steels for core internals in PWR. For comparison, the works on IASCC in BWR were also reviewed and briefly introduced in this report.

  15. 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

  16. 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

  17. Design of a PWR emergency core cooling simulator loop

    International Nuclear Information System (INIS)

    Melo, C.A. de.

    1982-12-01

    The preliminary design of a PWR Emergency Core Cooling Simulator Loop for investigations of the phenomena involved in a postulated Loss-of-Coolant Accident, during the Reflooding Phase, is presented. The functions of each component of the loop, the design methods and calculations, the specification of the instrumentation, the system operation sequence, the materials list and a cost assessment are included. (Author) [pt

  18. CORD, PWR Core Design and Fuel Management

    International Nuclear Information System (INIS)

    Trkov, Andrej

    1996-01-01

    1 - Description of program or function: CORD-2 is intended for core design applications of pressurised water reactors. The main objective was to assemble a core design system which could be used for simple calculations (such as frequently required for fuel management) as well as for accurate calculations (for example, core design after refuelling). 2 - Method of solution: The calculations are performed at the cell level with a lattice code in the supercell approximation to generate the single cell cross sections. Fuel assembly cross section homogenization is done in the diffusion approximation. Global core calculations can be done in the full three-dimensional cartesian geometry. Thermohydraulic feedbacks can be accounted for. The Effective Diffusion Homogenization method is used for generating the homogenized cross sections. 3 - Restrictions on the complexity of the problem: The complexity of the problem is selected by the user, depending on the capacity of his computer

  19. Core catcher concepts future PWR-Plants

    International Nuclear Information System (INIS)

    Alsmeyer, H.; Werle, H.

    1994-01-01

    Light water reactors of the next generation should have still greater passive safety, even in the most serious accidents. This includes the long term safe inclusion of the core inventory in the case of core meltdown accidents. The three concepts for cooling the liquefied core outside the reactor pressure vessel examined by KfK should remove the post-shutdown heat by direct contact of the melt with water. The geometric distribution of the melt increases its surface area, so that favourable conditions for heat removal from the poorly thermally-conducting melt are created and complete quick solidification occurs. The experiments examine both the relocation and distribution mechanisms of the melt and the reactions occurring when water enters. As strong interaction is possible on direct contact of the melt with water, an important aim is experimental determination and limitation of any resulting mechanical stresses. (orig./HP) [de

  20. Core power capability verification for PWR NPP

    International Nuclear Information System (INIS)

    Xian Chunyu; Liu Changwen; Zhang Hong; Liang Wei

    2002-01-01

    The Principle and methodology of pressurized water reactor nuclear power plant core power capability verification for reload are introduced. The radial and axial power distributions of normal operation (category I or condition I) and abnormal operation (category II or condition II) are simulated by using neutronics calculation code. The linear power density margin and DNBR margin for both categories, which reflect core safety, are analyzed from the point view of reactor physics and T/H, and thus category I operating domain and category II protection set point are verified. Besides, the verification results of reference NPP are also given

  1. Reverse depletion method for PWR core reload design

    International Nuclear Information System (INIS)

    Downar, T.J.; Kim, Y.J.

    1985-01-01

    Low-leakage fuel management is currently practiced in over half of all pressurized water reactor (PWR) cores. Prospects for even greater use of in-board fresh fuel loading are good as utilities seek to reduce core vessel fluence, mitigate pressurized thermal shock concerns, and extend vessel lifetime. Consequently, large numbers of burnable poison (BP) pins are being used to control the power peaking at the in-board fresh fuel positions. This has presented an additional complexity to the core reload design problem. In addition to determining the best location of each assembly in the core, the designer must concurrently determine the distribution of BP pins in the fresh fuel. A procedure was developed that utilizes the well-known Haling depletion to achieve an end-of-cycle (EOC) core state where the assembly pattern is configured in the absence of all control poison. This effectively separates the assembly assignment and BP distribution problems. Once an acceptable pattern at EOC is configured, the burnable and soluble poison required to control the power and core excess reactivity are solved for as unknown variables while depleting the cycle in reverse from the EOC exposure distribution to the beginning of cycle. The methods developed were implemented in an approved light water reactor licensing code to ensure the validity of the results obtained and provide for the maximum utility to PWR core reload design

  2. Three dimensions transport calculations for PWR core

    International Nuclear Information System (INIS)

    Richebois, E.

    2000-01-01

    The objective of this work is to define improved 3-D core calculation methods based on the transport theory. These methods can be particularly useful and lead to more precise computations in areas of the core where anisotropy and steep flux gradients occur, especially near interface and boundary conditions and in regions of high heterogeneity (bundle with absorbent rods). In order to apply the transport theory a new method for calculating reflector constants has been developed, since traditional methods were only suited for 2-group diffusion core calculations and could not be extrapolated to transport calculations. In this thesis work, the new method for obtaining reflector constants is derived regardless of the number of energy groups and of the operator used. The core calculations results using the reflector constants thereof obtained have been validated on the EDF's power reactor Saint Laurent B1 with MOX loading. The advantages of a 3-D core transport calculation scheme have been highlighted as opposed to diffusion methods; there are a considerable number of significant effects and potential advantages to be gained in rod worth calculations for instance. These preliminary results obtained with on particular cycle will have to be confirmed by more systematic analysis. Accidents like MSLB (main steam line break) and LOCA (loss of coolant accident) should also be investigated and constitute challenging situations where anisotropy is high and/or flux gradients are steep. This method is now being validated for others EDF's PWRs' reactors, as well as for experimental reactors and other types of commercial reactors. (author)

  3. Core management and performance analysis for PWR

    International Nuclear Information System (INIS)

    Lee, J.B.; Lee, C.K.; Kim, J.S.; Lee, S.K.; Moon, K.S.; Chun, B.J.; Chang, J.W.; Kim, Y.J.

    1981-01-01

    The KINS (KAERI Improved Nodal Simulation) program, a three-dimensional nodal simulation code for pressurized water reactor fuel management, has been developed and benchmarked against the cycles 1 and 2 of the Kori-1 reactor. The critical boron concentration and three-dimensional power distribution at BOL, HZP condition have been calculated and compared with the operating data. A three-dimensional depletion calculation at HFP condition has been performed for cycle 1 with an interval of 1000 MWD/MTU and compared with the operating data. Similar calculation was also performed for cycle 2 and then compared with the design data of the reactor vendor. At the same time, a prediction of in-core detectors reaction rate was made so as to be compared with the operating data. As the result of comparisons, our calculation as well as the justification of the correlations is shown to be in excellent agreement with the operating data within an allowable limit

  4. A nodal model for the simulation of a PWR core

    International Nuclear Information System (INIS)

    Souza Pinto, R. de.

    1981-06-01

    A computer program FORTRAN language was developed to simulate the neutronic and thermal-hydraulic transient behaviour of a PWR reactor core. The reator power is calculated using a point kinectics model with six groups of delayed neutron precursors. The fission product decay heat was considered assuming three effective decay heat groups. A nodal model was employed for the treatment of heat transfer in the fuel rod, with integration of the heat equation by the lumped parameter technique. Axial conduction was neglected. A single-channel nodal model was developed for the thermo-hydrodynamic simulation using mass and energy conservation equations for the control volumes. The effect of the axial pressure variation was neglected. The computer program was tested, with good results, through the simulation of the transient behaviour of postulated accidents in a typical PWR. (Author) [pt

  5. Thermal-hydraulic analysis of PWR cores in transient condition

    International Nuclear Information System (INIS)

    Silva Galetti, M.R. da.

    1984-01-01

    A calculational methodology for thermal - hydraulic analysis of PWR cores under steady-state and transient condition was selected and made available to users. An evaluation of the COBRA-IIIP/MIT code, used for subchannel analysis, was done through comparison of the code results with experimental data on steady state and transient conditions. As a result, a comparison study allowing spatial and temporal localization of critical heat flux was obtained. A sensitivity study of the simulation model to variations in some empirically determined parameter is also presented. Two transient cases from Angra I FSAR were analysed, showing the evolution of minimum DNBR with time. (Author) [pt

  6. Four-fluid model of PWR degraded cores

    International Nuclear Information System (INIS)

    Dearing, J.F.

    1985-01-01

    This paper describes the new two-dimensional, four-fluid fluid dynamics and heat transfer (FLUIDS) module of the MELPROG code. MELPROG is designed to give an integrated, mechanistic treatment of pressurized water reactor (PWR) core meltdown accidents from accident initiation to vessel melt-through. The code has a modular data storage and transfer structure, with each module providing the others with boundary conditions at each computational time step. Thus the FLUIDS module receives mass and energy source terms from the fuel pin module, the structures module, and the debris bed module, and radiation energy source terms from the radiation module. MELPROG, which models the reactor vessel, is also designed to model the vessel as a component in the TRAC/PF1 networking solution of a PWR reactor coolant system (RCS). The coupling between TRAC and MELPROG is implicit in the fluid dynamics of the reactor coolant (liquid water and steam) allowing an accurate simulation of the coupling between the vessel and the rest of the RCS during an accident. This paper deals specifically with the numerical model of fluid dynamics and heat transfer within the reactor vessel, which allows a much more realistic simulation (with less restrictive assumptions on physical behavior) of the accident than has been possible before

  7. Influence of spectral history on PWR full core calculation results

    International Nuclear Information System (INIS)

    Bilodid, Y.; Mittag, S.

    2011-01-01

    The few-group cross section libraries, used by reactor dynamics codes, are affected by the spectral history effect-a dependence of fuel cross sections not only on burnup, but also on local spectral conditions during burnup. A cross section correction method based on Pu-239 concentration was implemented in the reactor dynamic code DYN3D. This paper describes the influence of a cross section correction on full-core calculation results. Steady-state and burnup characteristics of a PWR equilibrium cycle, calculated by DYN3D with and without cross section corrections, are compared. A study has shown a significant influence of spectral history on axial power and burnup distributions as well as on calculated cycle length. An impact of the correction on transient calculations is studied for a control rod ejection example. (Authors)

  8. Construction and utilization of linear empirical core models for PWR in-core fuel management

    International Nuclear Information System (INIS)

    Okafor, K.C.

    1988-01-01

    An empirical core-model construction procedure for pressurized water reactor (PWR) in-core fuel management is developed that allows determining the optimal BOC k ∞ profiles in PWRs as a single linear-programming problem and thus facilitates the overall optimization process for in-core fuel management due to algorithmic simplification and reduction in computation time. The optimal profile is defined as one that maximizes cycle burnup. The model construction scheme treats the fuel-assembly power fractions, burnup, and leakage as state variables and BOC zone enrichments as control variables. The core model consists of linear correlations between the state and control variables that describe fuel-assembly behavior in time and space. These correlations are obtained through time-dependent two-dimensional core simulations. The core model incorporates the effects of composition changes in all the enrichment control zones on a given fuel assembly and is valid at all times during the cycle for a given range of control variables. No assumption is made on the geometry of the control zones. A scatter-composition distribution, as well as annular, can be considered for model construction. The application of the methodology to a typical PWR core indicates good agreement between the model and exact simulation results

  9. Benefits of Low Boron Core Design Concept for PWR

    Energy Technology Data Exchange (ETDEWEB)

    Daing, Aung Tharn; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

    2009-10-15

    Nuclear design study was carried out to develop low boron core (LBC) based on one of current PWR concepts, OPR-1000. Most of design parameters were the same with those of Ulchin unit-5 except extensive utilization of burnable poison (BP) pins in order to compensate reactivity increase in LBC. For replacement of reduced soluble boron concentration, four different kinds of integral burnable absorbers (IBAs) such as gadolinia, integral fuel burnable absorber (IFBA), erbia and alumina boron carbide were considered in suppressing more excess reactivity. A parametric study was done to find the optimal core options from many design candidates for fuel assemblies and cores. Among them, the most feasible core design candidate was chosen in accordance with general design requirements. In this paper, the feasibility and design change benefits of the most favorable LBC design were investigated in more detail through the comparison of neutronic and thermal hydraulic design parameters of LBC with the reference plant (REF). As calculation tools, the HELIOS/MASTER code package and the MATRA code were utilized. The main purpose of research herein is to estimate feasibility and capability of LBC which was mainly designed to mitigate boron dilution accident (BDA), and for reduction of corrosion products. The LBC design concept using lower boron concentration with an elevated enrichment in {sup 10}B allows a reduction in the concentration of lithium in the primary coolant required to maintain the optimum coolant pH. All in all, LBC with operation at optimum pH is expected to achieve some benefits from radiation source reduction of reduced corrosion product, the limitation of the Axial Offset Anomaly (AOA) and fuel cladding corrosion. Additionally, several merits of LBC are closely related to fluid systems and system related aspects, reduced boron and lithium costs, equipment size reduction for boric acid systems, elimination of heat tracing, and more aggressive fuel design concepts.

  10. Benefits of Low Boron Core Design Concept for PWR

    International Nuclear Information System (INIS)

    Daing, Aung Tharn; Kim, Myung Hyun

    2009-01-01

    Nuclear design study was carried out to develop low boron core (LBC) based on one of current PWR concepts, OPR-1000. Most of design parameters were the same with those of Ulchin unit-5 except extensive utilization of burnable poison (BP) pins in order to compensate reactivity increase in LBC. For replacement of reduced soluble boron concentration, four different kinds of integral burnable absorbers (IBAs) such as gadolinia, integral fuel burnable absorber (IFBA), erbia and alumina boron carbide were considered in suppressing more excess reactivity. A parametric study was done to find the optimal core options from many design candidates for fuel assemblies and cores. Among them, the most feasible core design candidate was chosen in accordance with general design requirements. In this paper, the feasibility and design change benefits of the most favorable LBC design were investigated in more detail through the comparison of neutronic and thermal hydraulic design parameters of LBC with the reference plant (REF). As calculation tools, the HELIOS/MASTER code package and the MATRA code were utilized. The main purpose of research herein is to estimate feasibility and capability of LBC which was mainly designed to mitigate boron dilution accident (BDA), and for reduction of corrosion products. The LBC design concept using lower boron concentration with an elevated enrichment in 10 B allows a reduction in the concentration of lithium in the primary coolant required to maintain the optimum coolant pH. All in all, LBC with operation at optimum pH is expected to achieve some benefits from radiation source reduction of reduced corrosion product, the limitation of the Axial Offset Anomaly (AOA) and fuel cladding corrosion. Additionally, several merits of LBC are closely related to fluid systems and system related aspects, reduced boron and lithium costs, equipment size reduction for boric acid systems, elimination of heat tracing, and more aggressive fuel design concepts

  11. Continuous firefly algorithm applied to PWR core pattern enhancement

    Energy Technology Data Exchange (ETDEWEB)

    Poursalehi, N., E-mail: npsalehi@yahoo.com [Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Tehran (Iran, Islamic Republic of); Zolfaghari, A.; Minuchehr, A.; Moghaddam, H.K. [Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Tehran (Iran, Islamic Republic of)

    2013-05-15

    Highlights: ► Numerical results indicate the reliability of CFA for the nuclear reactor LPO. ► The major advantages of CFA are its light computational cost and fast convergence. ► Our experiments demonstrate the ability of CFA to obtain the near optimal loading pattern. -- Abstract: In this research, the new meta-heuristic optimization strategy, firefly algorithm, is developed for the nuclear reactor loading pattern optimization problem. Two main goals in reactor core fuel management optimization are maximizing the core multiplication factor (K{sub eff}) in order to extract the maximum cycle energy and minimizing the power peaking factor due to safety constraints. In this work, we define a multi-objective fitness function according to above goals for the core fuel arrangement enhancement. In order to evaluate and demonstrate the ability of continuous firefly algorithm (CFA) to find the near optimal loading pattern, we developed CFA nodal expansion code (CFANEC) for the fuel management operation. This code consists of two main modules including CFA optimization program and a developed core analysis code implementing nodal expansion method to calculate with coarse meshes by dimensions of fuel assemblies. At first, CFA is applied for the Foxholes test case with continuous variables in order to validate CFA and then for KWU PWR using a decoding strategy for discrete variables. Results indicate the efficiency and relatively fast convergence of CFA in obtaining near optimal loading pattern with respect to considered fitness function. At last, our experience with the CFA confirms that the CFA is easy to implement and reliable.

  12. Continuous firefly algorithm applied to PWR core pattern enhancement

    International Nuclear Information System (INIS)

    Poursalehi, N.; Zolfaghari, A.; Minuchehr, A.; Moghaddam, H.K.

    2013-01-01

    Highlights: ► Numerical results indicate the reliability of CFA for the nuclear reactor LPO. ► The major advantages of CFA are its light computational cost and fast convergence. ► Our experiments demonstrate the ability of CFA to obtain the near optimal loading pattern. -- Abstract: In this research, the new meta-heuristic optimization strategy, firefly algorithm, is developed for the nuclear reactor loading pattern optimization problem. Two main goals in reactor core fuel management optimization are maximizing the core multiplication factor (K eff ) in order to extract the maximum cycle energy and minimizing the power peaking factor due to safety constraints. In this work, we define a multi-objective fitness function according to above goals for the core fuel arrangement enhancement. In order to evaluate and demonstrate the ability of continuous firefly algorithm (CFA) to find the near optimal loading pattern, we developed CFA nodal expansion code (CFANEC) for the fuel management operation. This code consists of two main modules including CFA optimization program and a developed core analysis code implementing nodal expansion method to calculate with coarse meshes by dimensions of fuel assemblies. At first, CFA is applied for the Foxholes test case with continuous variables in order to validate CFA and then for KWU PWR using a decoding strategy for discrete variables. Results indicate the efficiency and relatively fast convergence of CFA in obtaining near optimal loading pattern with respect to considered fitness function. At last, our experience with the CFA confirms that the CFA is easy to implement and reliable

  13. Recycling schemes of Americium targets in PWR/MOX cores

    International Nuclear Information System (INIS)

    Maldague, Th.; Pilate, S.; Renard, A.; Harislur, A.; Mouney, H.; Rome, M.

    1999-01-01

    From the orientation studies performed so far, both ways to recycle Am in PWR/MOX cores, homogeneous in MOX or heterogeneous in target pins, appear feasible, provided that enriched UO 2 is used as support of the MOX fuel. Multiple recycling can then proceed and stabilize Pu and Am quantities. With respect to the Pu multiple recycling strategy, recycling Am in addition needs 1/3 more 235 U, and creates 3 times more Curium. Thus, although feasible, such a fuel cycle is complicated and brings about a significant cost penalty, not quantified yet. The advantage of the heterogeneous option is to allow to manage in different ways the Pu in MOX fuel and the Am in target pins. For example, should Am remain combined to Cm after reprocessing, the recycling of a mix of Am+Cm could be deferred to let Cm transform into Pu before irradiation. The Am+Cm targets could also stay longer in the reactor, so as to avoid further reprocessing if possible. (author)

  14. Advanced methods for the study of PWR cores

    International Nuclear Information System (INIS)

    Lambert, M.; Salvatores, St.; Ferrier, A.; Pelet, J.; Nicaise, N.; Pouliquen, J.Y.; Foret, F.; Chauliac, C.; Johner, J.; Cohen, Ch.

    2003-01-01

    This document gathers the transparencies presented at the 6. technical session of the French nuclear energy society (SFEN) in October 2003. The transparencies of the annual meeting are presented in the introductive part: 1 - status of the French nuclear park: nuclear energy results, management of an exceptional climatic situation: the heat wave of summer 2003 and the power generation (J.C. Barral); 2 - status of the research on controlled thermonuclear fusion (J. Johner). Then follows the technical session about the advanced methods for the study of PWR reactor cores: 1 - the evolution approach of study methodologies (M. Lambert, J. Pelet); 2 - the point of view of the nuclear safety authority (D. Brenot); 3 - the improved decoupled methodology for the steam pipe rupture (S. Salvatores, J.Y. Pouliquen); 4 - the MIR method for the pellet-clad interaction (renovated IPG methodology) (E. Baud, C. Royere); 5 - the improved fuel management (IFM) studies for Koeberg (C. Cohen); 6 - principle of the methods of accident study implemented for the European pressurized reactor (EPR) (F. Foret, A. Ferrier); 7 - accident studies with the EPR, steam pipe rupture (N. Nicaise, S. Salvatores); 8 - the co-development platform, a new generation of software tools for the new methodologies (C. Chauliac). (J.S.)

  15. Feasibility of using gadolinium as a burnable poison in PWR cores. Final report

    International Nuclear Information System (INIS)

    Rothleder, B.M.

    1981-02-01

    As an alternative to the use of lumped burnable absorbers in PWR cores, distributed burnable absorbers are being considered for generic application. These burnable absorbers take the form of Gd 2 O 3 mixed with UO 2 in selected fuel rods (as is currently done in BWR cores). The work discussed herein concerns a three-dimensional feasibility study of the use of such distributed burnable absorbers in PWR cores. This study of distributed burnable absorbers was performed for the first cycle of a typical current design PWR using the following steps: analysis of a generic reference core design; determination of gadolinium assembly designs; determination of a generic gadolinium core design; evaluation of feasibility by examining selected parameters; and redesign of the generic gadolinium core, using axial zoning

  16. 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)

  17. Core losses of a permanent magnet synchronous motor with an amorphous stator core under inverter and sinusoidal excitations

    Science.gov (United States)

    Yao, Atsushi; Sugimoto, Takaya; Odawara, Shunya; Fujisaki, Keisuke

    2018-05-01

    We report core loss properties of permanent magnet synchronous motors (PMSM) with amorphous magnetic materials (AMM) core under inverter and sinusoidal excitations. To discuss the core loss properties of AMM core, a comparison with non-oriented (NO) core is also performed. In addition, based on both experiments and numerical simulations, we estimate higher (time and space) harmonic components of the core losses under inverter and sinusoidal excitations. The core losses of PMSM are reduced by about 59% using AMM stator core instead of NO core under sinusoidal excitation. We show that the average decrease obtained by using AMM instead of NO in the stator core is about 94% in time harmonic components.

  18. Pu recycling in a full Th-MOX PWR core. Part I: Steady state analysis

    International Nuclear Information System (INIS)

    Fridman, E.; Kliem, S.

    2011-01-01

    Research highlights: → Detailed 3D 100% Th-MOX PWR core design is developed. → Pu incineration increased by a factor of 2 as compared to a full MOX PWR core. → The core controllability under steady state conditions is demonstrated. - Abstract: Current practice of Pu recycling in existing Light Water Reactors (LWRs) in the form of U-Pu mixed oxide fuel (MOX) is not efficient due to continuous Pu production from U-238. The use of Th-Pu mixed oxide (TOX) fuel will considerably improve Pu consumption rates because virtually no new Pu is generated from thorium. In this study, the feasibility of Pu recycling in a typical pressurized water reactor (PWR) fully loaded with TOX fuel is investigated. Detailed 3-dimensional 100% TOX and 100% MOX PWR core designs are developed. The full MOX core is considered for comparison purposes. The design stages included determination of Pu loading required to achieve 18-month fuel cycle assuming three-batch fuel management scheme, selection of poison materials, development of the core loading pattern, optimization of burnable poison loadings, evaluation of critical boron concentration requirements, estimation of reactivity coefficients, core kinetic parameters, and shutdown margin. The performance of the MOX and TOX cores under steady-state condition and during selected reactivity initiated accidents (RIAs) is compared with that of the actual uranium oxide (UOX) PWR core. Part I of this paper describes the full TOX and MOX PWR core designs and reports the results of steady state analysis. The TOX core requires a slightly higher initial Pu loading than the MOX core to achieve the target fuel cycle length. However, the TOX core exhibits superior Pu incineration capabilities. The significantly degraded worth of control materials in Pu cores is partially addressed by the use of enriched soluble boron and B 4 C as a control rod absorbing material. Wet annular burnable absorber (WABA) rods are used to flatten radial power distribution

  19. Transient performance and design aspects of low boron PWR cores with increased utilization of burnable absorbers

    International Nuclear Information System (INIS)

    Papukchiev, Angel; Schaefer, Anselm

    2008-01-01

    In conventional pressurized water reactor (PWR) designs, soluble boron is used for reactivity control over core fuel cycle. As high boron concentrations have significant impact on reactivity feedback properties and core transient behaviour, design changes to reduce boron concentration in the reactor coolant are of general interest in view of improving PWR inherent safety. In order to assess the potential advantages of such strategies in current PWRs, two low boron core configurations based on fuel with increased utilization of gadolinium and erbium burnable absorbers have been developed. The new PWR designs permit to reduce the natural boron concentration in reactor coolant at begin of cycle to 518 (Gd) and 805 (Er) ppm. An innovative low boron core design methodology was implemented combining a simplified reactivity balance search procedure with a core design approach based on detailed 3D diffusion calculations. Fuel cross sections needed for nuclear libraries were generated using the 2D lattice code HELIOS [2] and full core configurations were modelled with the 3D diffusion code QUABOX/CUBBOX [3]. For dynamic 3D calculations, the coupled code system ATHLET - QUABOX/CUBBOX was used [4]. The new cores meet German acceptance criteria regarding stuck rod, departure from nucleate boiling ratio (DNBR), shutdown margin, and maximal linear power. For the assessment of potential safety advantages of the new cores, comparative analyses were performed for three PWR core designs: the already mentioned two low boron designs and a standard design. The improved safety performance of the low boron cores in anticipated transients without scram (ATWS), boron dilution scenarios and beyond design basis accidents (BDBA) has already been reported in [1, 2 and 3]. This paper gives a short reminder on the results obtained. Moreover, it deals not only with the potential advantages, but also addresses the drawbacks of the new PWR configurations - complex core design, increased power

  20. Coolant flow monitoring in a PWR core using noise analysis

    International Nuclear Information System (INIS)

    Kostic, Lj.

    1992-01-01

    Experimental investigations of the neutron and temperature noise field have been performed in the 1350 MW PWR nuclear power plant. Evaluation in the low frequency range, where both feedback effects and different thermohydraulics phenomena are dominant, succeeded in measuring the coolant velocity. This is important for determination and localization of essential deviations and possible anomalies. (author)

  1. Neutronic feasibility of PWR core with mixed oxide fuels in the Republic of Korea

    International Nuclear Information System (INIS)

    Kim, Y.J.; Joo, H.K.; Jung, H.G.; Sohn, D.S.

    1997-01-01

    Neutronic feasibility of a PWR core with mixed oxide (MOX) fuels has been investigated as part of the feasibility study for recycling spent fuels in Korea. A typical 3-loop PWR with 900 MWe capacity is selected as reference plant to develop equilibrium core designs with low-leakage fuel management scheme, while incorporating various MOX loading. The fuel management analyses and limited safety analyses show that, safely stated, MOX recycling with 1/3 reload fraction can be accommodated for both annual and 18 month fuel cycle schemes in Korean PWRs, without major design modifications on the reactor systems. (author). 12 refs, 4 figs, 3 tabs

  2. A Feasibility Study on Core Cooling of Reduced-Moderation PWR for the Large Break LOCA

    International Nuclear Information System (INIS)

    Hiroyuki Yoshida; Akira Ohnuki; Hajime Akimoto

    2002-01-01

    A design study of a reduced-moderation water reactor (RMWR) with tight lattice core is being carried out at the Japan Atomic Energy Research Institute (JAERI) as one candidate for future reactors. The concept is developed to achieve a conversion ratio greater than unity using the tight lattice core (volume ratio of moderator to fuel is around 0.5 and the gap spacing between the fuel rods is remarkably narrower than in a reactor currently operated). Under such tight configuration, the core thermal margin becomes smaller and should be evaluated in a normal operation and also during the reflood phase in a large break loss-of-coolant accident (LBLOCA) for PWR type reactors. In this study, we have performed a feasibility evaluation on core cooling of reduced moderation PWR for the LBLOCA (200% break). The evaluation was performed for the primary system after the break by the REFLA/TRAC code. The core thermal output of the reduced moderation PWR is 2900 MWt, the gap between adjacent fuel rods is 1 mm, and heavy water is used as the moderator and coolant. The present design adopts seed fuel assemblies (MOX fuel) and several blanket fuel assemblies. In the blanket fuel assemblies, power density is lower than that of the seed fuel assemblies. Then, we set a channel box to each fuel assembly in order to adjust the flow rate in each assembly, because the possibility that the coolant boils in the seed fuel assemblies is very high. The pressure vessel diameter is bigger in comparison with a current PWR and core height is smaller than the current one. The current 4-loop PWR system is used, and, however, to fit into the bigger pressure vessel volume (about 1.5 times), we set up the capacity of the accumulator (1.5 times of the current PWR). Although the maximum clad temperature reached at about 1200 K in the position of 0.6 m from the lower core support plate, it is sufficiently lower than the design criteria of the current PWR (1500 K). The core cooling of the reduced moderation

  3. PWR core follow calculations using the ELCOS code system

    International Nuclear Information System (INIS)

    Grimm, P.; Paratte, J.M.

    1990-01-01

    The ELCOS code system developed at PSI is used to simulate a cycle of a PWR in which one fifth of the assemblies are MOX fuel. The reactor and the calculational methods are briefly described. The calculated critical boron concentrations and power distributions are compared with the measurements at the plant. Although the critical boron concentration is somewhat overpredicted and the computed power distributions are slightly flatter than the measured ones the results of the calculations agree generally well with the measured data. (author) 1 tab., 8 figs., 6 refs

  4. Study on Reactor Physics Characteristic of the PWR Core Using UO2

    International Nuclear Information System (INIS)

    Tukiran Surbakti

    2009-01-01

    Study on reactor physics characteristic of the PWR core using UO 2 fuel it is necessary to be done to know the characteristic of geometry, condition and configuration of pin cell in the fuel assembly Because the geometry, configuration and condition of the pin cell in fuel core determine the loading strategy of in-core fuel management Calculation of k e ff is a part of the neutronic core parameter calculation to know the reactor physics characteristic. Generally, core calculation is done using computer code starts from modelling one unit fuel lattice cell, fuel assembly, reflector, irradiation facility and until core reactor. In this research, the modelling of pin cell and fuel assembly of the PWR 17 ×17 is done homogeneously. Calculation of the k-eff is done with variation of the fuel volume fraction, fuel pin diameter, fuel enrichment. The calculation is using by NITAWL and CENTRM, and then the results will be compared to KENOVI code. The result showed that the value of k e ff for pin cell and fuel assembly PWR 17 ×17 is not different significantly with homogenous and heterogenous models. The results for fuel volume fraction of 0.5; rod pitch 1.26 cm and fuel pin diameter of 9.6 mm is critical with burn up of 35,0 GWd/t. The modeling and calculation method accurately is needed to calculation the core physic parameter, but sometimes, it is needed along time to calculate one model. (author)

  5. Core losses of a permanent magnet synchronous motor with an amorphous stator core under inverter and sinusoidal excitations

    Directory of Open Access Journals (Sweden)

    Atsushi Yao

    2018-05-01

    Full Text Available We report core loss properties of permanent magnet synchronous motors (PMSM with amorphous magnetic materials (AMM core under inverter and sinusoidal excitations. To discuss the core loss properties of AMM core, a comparison with non-oriented (NO core is also performed. In addition, based on both experiments and numerical simulations, we estimate higher (time and space harmonic components of the core losses under inverter and sinusoidal excitations. The core losses of PMSM are reduced by about 59% using AMM stator core instead of NO core under sinusoidal excitation. We show that the average decrease obtained by using AMM instead of NO in the stator core is about 94% in time harmonic components.

  6. 3-D full core calculations for the long-term behaviour of PWR's

    International Nuclear Information System (INIS)

    Winter, H.J.; Koebke, K.; Wagner, M.R.

    1987-01-01

    Presently, the most realistic simulation of a PWR core is by means of three-dimensional (3-D) full core calculations. Only by such 3-D representations can the large scope of axial effects be treated in an accurate and direct way, without the need to perform various auxiliary calculations. Although the computationally efficient burnup-corrected nodal expansion method (NEM-BC) is used, the computing effort for 3-D reactor calculations becomes rather high, e.g. a storage of about 320000 words is required to describe a 1300 MWe PWR. NEM-BC was introduced (1979) into KWU's package of PWR design codes because of its high accuracy and the great reduction of computing time and storage requirements in comparison to other methods. The application of NEM-BC to 3-dimensional PWR design is strongly correlated with the progress achieved in the solution of the homogenization and dehomogenization problem. By means of suitable methods (equivalence theory) the transport-theoretical information of the pinwise power and burnup distribution for the heterogeneous fuel assemblies is transferred in a consistent manner to the full core reactor solution. The new methods and the corresponding code system are explained in some detail. (orig.)

  7. A probabilistic SSYST-3 analysis for a PWR-core during a large break LOCA

    International Nuclear Information System (INIS)

    Schubert, J.D.; Gulden, W.; Jacobs, G.; Meyder, R.; Sengpiel, W.

    1985-05-01

    This report demonstrates the SSYST-3 analysis and application for a German PWR of 1300 MW. The report is concerned with the probabilistic analysis of a PWR core during a loss-of-coolant accident due to a large break. With the probabilistic analysis, the distribution functions of the maximum temperatures and cladding elongations occuring in the core can be calculated. Parameters like rod power, the thermohydraulic boundary conditions, stored energy in the fuel rods and the heat transfer coefficient were found to be the most important. The expected value of core damage was determined to be 2.9% on the base of response surfaces for cladding temperature and strain deduced from SSYST-3 single rod results. (orig./HP) [de

  8. Xenon oscillation tests in four-loop PWR cores

    International Nuclear Information System (INIS)

    Aoki, Norihiko; Osaka, Kenichi; Shimada, Shoichiro; Tochihara, Hiroshi; Machii, Seigo

    1980-01-01

    The Kansai Electric Power Co.'s OHI Unit 1 and 2 are the first 4-loop PWRs in Japan which use 17 x 17 fuel assemblies and have essentially the same plant parameters. A 4-loop core has larger core radius and higher power density than those of 2- or 3-loop cores, and is less stable for Xe oscillation. It is therefore important to confirm that Xe oscillations in radial direction are sufficiently stable in a 4-loop core. Radial and axial Xe oscillation tests were performed during the startup physics tests of OHI Unit 1 and 2; Xe oscillation was induced by perturbation of control rods and the Xe effect on power distribution observed periodically. The test results show that the transverse Xe oscillation in the 4-loop core is sufficiently stable and that the agreement between the measurement and the calculated prediction is good. (author)

  9. The development of emergency core cooling systems in the PWR, BWR, and HWR Candu type of nuclear power plants

    International Nuclear Information System (INIS)

    Mursid Djokolelono.

    1976-01-01

    Emergency core cooling systems in the PWR, BWR, and HWR-Candu type of nuclear power plant are reviewed. In PWR and BWR the emergency cooling can be catagorized as active high pressure, active low pressure, and a passive one. The PWR uses components of the shutdown cooling system: whereas the BWR uses components of pressure suppression contaiment. HWR Candu also uses the shutdown cooling system similar to the PWR except some details coming out from moderator coolant separation and expensive cost of heavy water. (author)

  10. Adaptive control of a PWR core power using neural networks

    International Nuclear Information System (INIS)

    Arab-Alibeik, H.; Setayeshi, S.

    2005-01-01

    Reactor power control is important because of safety concerns and the call for regular and appropriate operation of nuclear power plants. It seems that the load-follow operation of these plants will be unavoidable in the future. Discrepancies between the real plant and the model used in controller design for load-follow operation encourage one to use auto-tuning and (or) adaptive techniques. Neural network technology shows great promise for addressing many problems in non-model-based adaptive control methods. Also, there has been a great attention to inverse control especially in the neural and fuzzy control context. Fortunately, online adaptation eliminates some limitations of inverse control and its shortcomings for real world applications. We use a neural adaptive inverse controller to control the power of a PWR reactor. The stability of the system and convergence of the controller parameters are guaranteed during online adaptation phase provided the controller is near the plant's real inverse after offline training period. The performance of the controller is verified using nonlinear simulations in diverse operating conditions

  11. Optimal burnable poison utilization in PWR core reload design

    International Nuclear Information System (INIS)

    Downar, T.J.

    1986-01-01

    A method was developed for determining the optimal distribution and depletion of burnable poisons in a Pressurized Water Reactor core. The well-known Haling depletion technique is used to achieve the end-of-cycle core state where the fuel assembly arrangement is configured in the absence of all control poison. The soluble and burnable poison required to control the core reactivity and power distribution are solved for as unknown variables while step depleting the cycle in reverse with a target power distribution. The method was implemented in the NRC approved licensing code SIMULATE

  12. Transient analysis for PWR reactor core using neural networks predictors

    International Nuclear Information System (INIS)

    Gueray, B.S.

    2001-01-01

    In this study, transient analysis for a Pressurized Water Reactor core has been performed. A lumped parameter approximation is preferred for that purpose, to describe the reactor core together with mechanism which play an important role in dynamic analysis. The dynamic behavior of the reactor core during transients is analyzed considering the transient initiating events, wich are an essential part of Safety Analysis Reports. several transients are simulated based on the employed core model. Simulation results are in accord the physical expectations. A neural network is developed to predict the future response of the reactor core, in advance. The neural network is trained using the simulation results of a number of representative transients. Structure of the neural network is optimized by proper selection of transfer functions for the neurons. Trained neural network is used to predict the future responses following an early observation of the changes in system variables. Estimated behaviour using the neural network is in good agreement with the simulation results for various for types of transients. Results of this study indicate that the designed neural network can be used as an estimator of the time dependent behavior of the reactor core under transient conditions

  13. Some factors affecting radiative heat transport in PWR cores

    International Nuclear Information System (INIS)

    Hall, A.N.

    1989-04-01

    This report discusses radiative heat transport in Pressurized Water Reactor cores, using simple models to illustrate basic features of the transport process. Heat transport by conduction and convection is ignored in order to focus attention on the restrictions on radiative heat transport imposed by the geometry of the heat emitting and absorbing structures. The importance of the spacing of the emitting and absorbing structures is emphasised. Steady state temperature distributions are found for models of cores which are uniformly heated by fission product decay. In all of the models, a steady state temperature distribution can only be obtained if the central core temperature is in excess of the melting point of UO 2 . It has recently been reported that the MIMAS computer code, which takes into account radiative heat transport, has been used to model the heat-up of the Three Mile Island-2 reactor core, and the computations indicate that the core could not have reached the melting point of UO 2 at any time or any place. We discuss this result in the light of the calculations presented in this paper. It appears that the predicted stabilisation of the core temperatures at ∼ 2200 0 C may be a consequence of the artificially large spacing between the radial rings employed in the MIMAS code, rather than a result of physical significance. (author)

  14. ASCOT-1, Thermohydraulics of Axisymmetric PWR Core with Homogeneous Flow During LOCA

    International Nuclear Information System (INIS)

    1978-01-01

    1 - Nature of the physical problem solved: ASCOT-1 is used to analyze the thermo-hydraulic behaviour in a PWR core during a loss-of-coolant accident. 2 - Method of solution: The core is assumed to be axisymmetric two-dimensional and the conservation laws are solved by the method of characteristics. For the temperature response of fuel in the annular regions into which the core is divided, the heat conduction equations are solved by an explicit method with averaged flow conditions. 3 - Restrictions on the complexity of the problem: Axisymmetric two-dimensional homogeneous flows

  15. An expert system for PWR core operation management

    Energy Technology Data Exchange (ETDEWEB)

    Ida, Toshio; Masuda, Masahiro; Nishioka, Hiromasa

    1988-01-01

    Planning for restartup after planned or unplanned reactor shutdown and load-follow operations is an important task in the core operation management of pressurized water reactors (PWRs). These planning problems have been solved by planning experts using their expertise and the computational prediction of core behavior. Therefore, the quality of the plan and the time consumed in the planning depend heavily on the skillfulness of the planning experts. A knowledge engineering approach has been recently considered as a promising means to solve such complicated planning problems. Many knowledge-based systems have been developed so far, and some of them have already been applied because of their effectiveness. The expert system REPLEX has been developed to aid core management engineers in making a successful plan for the restartup or the load-follow operation of PWRs within a shorter time. It can maintain planning tasks at a high-quality level independent of the skillfulness of core management engineers and enhance the efficiency of management. REPLEX has an explanation function that helps user understanding of plans. It could be a useful took, therefore, for the training of core management engineers.

  16. An expert system for PWR core operation management

    International Nuclear Information System (INIS)

    Ida, Toshio; Masuda, Masahiro; Nishioka, Hiromasa.

    1988-01-01

    Planning for restartup after planned or unplanned reactor shutdown and load-follow operations is an important task in the core operation management of pressurized water reactors (PWRs). These planning problems have been solved by planning experts using their expertise and the computational prediction of core behavior. Therefore, the quality of the plan and the time consumed in the planning depend heavily on the skillfulness of the planning experts. A knowledge engineering approach has been recently considered as a promising means to solve such complicated planning problems. Many knowledge-based systems have been developed so far, and some of them have already been applied because of their effectiveness. The expert system REPLEX has been developed to aid core management engineers in making a successful plan for the restartup or the load-follow operation of PWRs within a shorter time. It can maintain planning tasks at a high-quality level independent of the skillfulness of core management engineers and enhance the efficiency of management. REPLEX has an explanation function that helps user understanding of plans. It could be a useful took, therefore, for the training of core management engineers

  17. 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

  18. Preliminary analysis of a large 1600 MWe PWR core loaded with 30% MOX fuel

    International Nuclear Information System (INIS)

    Polidoro, Franco; Corsetti, Edoardo; Vimercati, Giuliano

    2011-01-01

    The paper presents a full-core 3-D analysis of the performances of a large 1600 MWe PWR core, loaded with 30% MOX fuel, in accordance with the European Utility Requirements (EUR). These requirements state that the European next generation power plants have to be designed capable to use MOX (UO 2 - PuO 2 ) fuel assemblies up to 50% of the core, together with UO 2 fuel assemblies. The use of MOX assemblies has a significant impact on key physic parameters and on safety. A lot of studies have been carried out in the past to explore the feasibility of plutonium recycling strategies by loading LWR reactors with MOX fuel. Many of these works were based on lattice codes, in order to perform detailed analyses of the neutronic characteristics of MOX assemblies. With the aim to take into account their interaction with surrounding UO 2 fuel elements, and the global effects on the core at operational conditions, an integrated approach making use of a 3-D core simulation is required. In this light, the present study adopts the state-of-art numerical models CASMO-5 and SIMULATE-3 to analyze the behavior of the core fueled with 30% MOX and to compare it with that of a large PWR reference core, fueled with UO 2 . (author)

  19. Fluid-structure coupled dynamic response of PWR core barrel during LOCA

    International Nuclear Information System (INIS)

    Lu, M.W.; Zhang, Y.G.; Shi, F.

    1991-01-01

    This paper is engaged in the Fluid-Structure Interaction LOCA analysis of the core barrel of PWR. The analysis is performed by a multipurpose computer code SANES. The FSI inside the pressure vessel is treated by a FEM code including some structural and acoustic elements. The transient in the primary loop is solved by a two-phase flow code. Both codes are coupled one another. Some interesting conclusions are drawn. (author)

  20. Generalized perturbation theory error control within PWR core-loading pattern optimization

    International Nuclear Information System (INIS)

    Imbriani, J.S.; Turinsky, P.J.; Kropaczek, D.J.

    1995-01-01

    The fuel management optimization code FORMOSA-P has been developed to determine the family of near-optimum loading patterns for PWR reactors. The code couples the optimization technique of simulated annealing (SA) with a generalized perturbation theory (GPT) model for evaluating core physics characteristics. To ensure the accuracy of the GPT predictions, as well as to maximize the efficient of the SA search, a GPT error control method has been developed

  1. Natural vibrations of a core banel of a PWR type reactor by elements of revolution shell

    International Nuclear Information System (INIS)

    Barcellos, C.S. de.

    1980-01-01

    Aim to estimate the behavior of the cove barrel of PWR type reactors, submitted to several load conditions, their dynamic characteristic, were determined. In order to obtain the natural modes and frequencies of the core barrel, the CYLDYFE comprete code based in the finite element method, was developed. The obtained results are compared with results obtained by other programs such as SAP, ASKA and STRUDL/DYNAL and by other analytical methods. (M.C.K.) [pt

  2. Evaluation of full MOX core capability for a 900 MWe PWR

    International Nuclear Information System (INIS)

    Joo, Hyung-Kook; Kim, Young-Jin; Jung, Hyung-Guk; Kim, Young-Il; Sohn, Dong-Seong

    1996-01-01

    Full MOX capability of a PWR core with 900 MWe capacity has been evaluated in view of plutonium consumption and design feasibility as an effective means for spent fuel management. Three full MOX cores have been conceptually designed; for annual cycle, for 18-month cycle, and for 18-month cycle with high moderation lattice. Fissile and total plutonium quantities at discharge are significantly reduced to 60% and 70% respectively of initial value for standard full MOX cores. It is estimated that one full MOX core demands about 1 tonne of plutonium per year to be reloaded, which is equivalent to reprocessing of spent nuclear fuels discharged from five nuclear reactors operating with uranium fuels. With low-leakage loading scheme, a full MOX core with either annual or 18-month cycle can be designed satisfactorily by installing additional rod cluster control system and modifying soluble boron system. Overall high moderation lattice case promises better core nuclear characteristics. (author)

  3. Feasibility study on thermal-hydraulic design of reduced-moderation PWR-type core

    International Nuclear Information System (INIS)

    Yoshida, Hiroyuki; Ohnuki, Akira; Akimoto, Hajime

    2000-03-01

    At JAERI, a conceptual study on reduced-moderation water reactor (RMWR) has been performed as one of the advanced reactor system which is designed so as to realize the conversion ratio more than unity. In this reactor concept, the gap spacing between the fuel rods is remarkably narrower than in a reactor currently operated. Therefore, an evaluation of the core thermal margin becomes very important in the design of the RMWR. In this study, we have performed a feasibility evaluation on thermal-hydraulic design of RM-PWR type core (core thermal output: 2900 MWt, Rod gaps: 1 mm). In RM-PWR core, seed and blanket regions are exist. In the blanket region, power density is lower than that of the seed region. Then, evaluation was performed under setting a channel box to each fuel assembly in order to adjust the flow rate in each assembly, because it is possible that the coolant boils in the seed region. In the feasibility evaluations, subchannel code COBRA-IV-I was used in combination with KfK DNB (departure nucleate boiling) correlation. When coolant mass flow rate to the blanket fuel assembly is reduced by 40%, and that to the seed fuel assembly is increased, coolant boiling is not occurred in the assembly region calculation. Provided that the channel boxes to the blanket fuel assembly are set up and coolant mass flow rate to the blanket fuel assembly is reduced by 40%, it is confirmed by the whole core calculation that the boiling of the coolant is not occurred and the RM-PWR core is feasible. (author)

  4. Study and analysis for the flow-induced vibration of the core barrel of a PWR

    International Nuclear Information System (INIS)

    Yao Weida; Shi Guolin; Jiang Nanyan

    1989-01-01

    The resemblance criteria are derived and a test model is designed by applying the flow-soild coupling theory. After having completed the model analysis of the pressurized water reactor (PWR) core barrel in an 1:10 model, the dynamic characteristics are obtained. In an 1:5 reactor model with a hydraulic closed loop, the hydraulic vibration tests of the core barrel are performed, and the relations between the flow rate and the flow-induced pulse pressure on core barrel, acceleration and strain signals have been measured. The corresponding responses and a group of computational equations for hydraulic vibration are derived from these two experiments. The computational hydraulic vibration responses for core barrel in Qinshan Nuclear Power Plant are in good agreement with the test results, and it shows that the core barrel is safe within its lifetime of 30 years

  5. Application of the pertubation theory to a two channels model for sensitivity calculations in PWR cores

    International Nuclear Information System (INIS)

    Oliveira, A.C.J.G. de; Andrade Lima, F.R. de

    1989-01-01

    The present work is an application of the perturbation theory (Matricial formalism) to a simplified two channels model, for sensitivity calculations in PWR cores. Expressions for some sensitivity coefficients of thermohydraulic interest were developed from the proposed model. The code CASNUR.FOR was written in FORTRAN to evaluate these sensitivity coefficients. The comparison between results obtained from the matrical formalism of pertubation theory with those obtained directly from the two channels model, makes evident the efficiency and potentiality of this perturbation method for nuclear reactor cores sensitivity calculations. (author) [pt

  6. Reactor core design calculations and fuel management in PWR; Izracun projekta sredice in upravljanja z forivom tlacnovodnega reaktorja

    Energy Technology Data Exchange (ETDEWEB)

    Ravnik, M [Institut Jozef Stefan, Ljubljana (Yugoslavia)

    1987-07-01

    Computer programs and methods developed at J. Stefan Institute for nuclear core design of Krsko NPP are treated. development, scope, verification and organisation of core design procedure are presented. The core design procedure is applicable to any NPP of PWR type. (author)

  7. PWR core safety analysis with 3-dimensional methods

    International Nuclear Information System (INIS)

    Gensler, A.; Kühnel, K.; Kuch, S.

    2015-01-01

    Highlights: • An overview of AREVA’s safety analysis codes their coupling is provided. • The validation base and licensing applications of these codes are summarized. • Coupled codes and methods provide improved margins and non-conservative results. • Examples for REA and inadvertent opening of the pressurizer safety valve are given. - Abstract: The main focus of safety analysis is to demonstrate the required safety level of the reactor core. Because of the demanding requirements, the quality of the safety analysis strongly affects the confidence in the operational safety of a reactor. To ensure the highest quality, it is essential that the methodology consists of appropriate analysis tools, an extensive validation base, and last but not least highly educated engineers applying the methodology. The sophisticated 3-dimensional core models applied by AREVA ensure that all physical effects relevant for safety are treated and the results are reliable and conservative. Presently AREVA employs SCIENCE, CASMO/NEMO and CASCADE-3D for pressurized water reactors. These codes are currently being consolidated into the next generation 3D code system ARCADIA®. AREVA continuously extends the validation base, including measurement campaigns in test facilities and comparisons of the predictions of steady state and transient measured data gathered from plants during many years of operation. Thus, the core models provide reliable and comprehensive results for a wide range of applications. For the application of these powerful tools, AREVA is taking benefit of its interdisciplinary know-how and international teamwork. Experienced engineers of different technical backgrounds are working together to ensure an appropriate interpretation of the calculation results, uncertainty analysis, along with continuously maintaining and enhancing the quality of the analysis methodologies. In this paper, an overview of AREVA’s broad application experience as well as the broad validation

  8. Assessment of void swelling in austenitic stainless steel PWR core internals

    International Nuclear Information System (INIS)

    Chung, H.M.

    2006-01-01

    As many pressurized water reactors (PWRs) age and life extension of the aged plants is considered, void swelling behavior of austenitic stainless steel (SS) core internals has become the subject of increasing attention. In this report, the available database on void swelling and density change of austenitic SSs was critically reviewed. Irradiation conditions, test procedures, and microstructural characteristics were carefully examined, and key factors that are important to determine the relevance of the database to PWR conditions were evaluated. Most swelling data were obtained from steels irradiated in fast breeder reactors at temperatures >385 C and at dose rates that are orders of magnitude higher than PWR dose rates. Even for a given irradiation temperature and given steel, the integral effects of dose and dose rate on void swelling should not be separated. It is incorrect to extrapolate swelling data on the basis of 'progressive compounded multiplication' of separate effects of factors such as dose, dose rate, temperature, steel composition, and fabrication procedure. Therefore, the fast reactor data should not be extrapolated to determine credible void swelling behavior for PWR end-of-life (EOL) or life-extension conditions. Although the void swelling data extracted from fast reactor studies is extensive and conclusive, only limited amounts of swelling data and information have been obtained on microstructural characteristics from discharged PWR internals or steels irradiated at temperatures and at dose rates comparable to those of a PWR. Based on this relatively small amount of information, swelling in thin-walled tubes and baffle bolts in a PWR is not considered a concern. As additional data and relevant research becomes available, the newer results should be integrated with existing data, and the worthiness of this conclusion should continue to be scrutinized. PWR baffle reentrant corners are the most likely location to experience high swelling rates, and

  9. A new uncertainty reduction method for PWR cores with erbia bearing fuel

    International Nuclear Information System (INIS)

    Takeda, Toshikazu; Sano, Tadafumi; Kitada, Takanori; Kuroishi, Takeshi; Yamasaki, Masatoshi; Unesaki, Hironobu

    2008-01-01

    The concept of a PWR with erbia bearing high burnup fuel has been proposed. The erbia is added to all fuel with over 5% 235 U enrichment to retain the neutronics characteristics to that within 5% 235 U enrichment. There is a problem of the prediction accuracy of the neutronics characteristics with erbia bearing fuel because of the short of experimental data of erbia bearing fuel. The purpose of the present work is to reduce the uncertainty. A new method has been proposed by combining the bias factor method and the cross section adjustment method. For the PWR core, the uncertainty reduction, which shows the rate of reduction of uncertainty, of the k eff is 0.865 by the present method and 0.801 by the conventional bias factor method. Thus the prediction uncertainties are reduced by the present method compared to the bias factor method. (authors)

  10. Computer code validation study of PWR core design system, CASMO-3/MASTER-α

    International Nuclear Information System (INIS)

    Lee, K. H.; Kim, M. H.; Woo, S. W.

    1999-01-01

    In this paper, the feasibility of CASMO-3/MASTER-α nuclear design system was investigated for commercial PWR core. Validation calculation was performed as follows. Firstly, the accuracy of cross section generation from table set using linear feedback model was estimated. Secondly, the results of CASMO-3/MASTER-α was compared with CASMO-3/NESTLE 5.02 for a few benchmark problems. Microscopic cross sections computed from table set were almost the same with those from CASMO-3. There were small differences between calculated results of two code systems. Thirdly, the repetition of CASMO-3/MASTER-α calculation for Younggwang Unit-3, Cycle-1 core was done and their results were compared with nuclear design report(NDR) and uncertainty analysis results of KAERI. It was found that uncertainty analysis results were reliable enough because results were agreed each other. It was concluded that the use of nuclear design system CASMO-3/MASTER-α was validated for commercial PWR core

  11. Advanced PWR Core Design with Siemens High-Plutonium-Content MOX Fuel Assemblies

    International Nuclear Information System (INIS)

    Dieter Porsch; Gerhard Schlosser; Hans-Dieter Berger

    2000-01-01

    The Siemens experience with plutonium recycling dates back to the late 1960s. Over the years, extensive research and development programs were performed for the qualification of mixed-oxide (MOX) technology and design methods. Today's typical reload enrichments for uranium and MOX fuel assemblies and modern core designs have become more demanding with respect to accuracy and reliability of design codes. This paper presents the status of plutonium recycling in operating high-burnup pressurized water reactor (PWR) cores. Based on actual examples, it describes the validation status of the design methods and stresses current and future needs for fuel assembly and core design including those related to the disposition of weapons-grade plutonium

  12. Flow with boiling in four-cusp channels simulating damaged core in PWR type reactors

    International Nuclear Information System (INIS)

    Esteves, M.M.

    1985-01-01

    The study of subcooled nucleate flow boiling in non-circular channels is of great importance to engineering applications in particular to Nuclear Engineering. In the present work, an experimental apparatus, consisting basically of a refrigeration system, running on refrigerant-12, has been developed. Preliminary tests were made with a circular tube. The main objective has been to analyse subcooled flow boiling in four-cusp channels simulating the flow conditions in a PWR core degraded by accident. Correlations were developed for the forced convection film coefficient for both single-phase and subcooled flow boiling. The incipience of boiling in such geometry has also been studied. (author) [pt

  13. STYCA, a computer program in the dynamic structural analysis of a PWR core

    International Nuclear Information System (INIS)

    Silva Macedo, L.V. da; Breyne Salvagni, R. de

    1992-01-01

    A procedure for the dynamic structural analysis of a PWR core is presented, impacts between fuel assemblies may occur because of the existence of gaps between them. Thus, the problem is non-linear and an spectral analysis is avoided. A time-history response analysis is necessary. The Modal Superposition Method with the Duhamel integral was used in order to solve the problem. An algorithm of solution and also results obtained with the STYCA computer program, developed on the basis of what was proposed here, are presented. (author)

  14. Experimental modelling of core debris dispersion from the vault under a PWR pressure vessel: Part 1

    International Nuclear Information System (INIS)

    Macbeth, R.V.; Trenberth, R.

    1987-12-01

    Modelling experiments have been done on a 1/25 scale model in Perspex of the vault under a PWR pressure vessel. Various liquids have been used to simulate molten core debris assumed to have fallen on to the vault floor from a breach at the bottom of the pressure vessel. High pressure air and helium have been used to simulate the discharge of steam and gas from the breach. The dispersion of liquid via the vault access shafts has been measured. Photographs have been taken of fluid flow patterns and velocity profiles have been obtained. The requirements for further experiments are indicated. (author)

  15. Development of the computer code system for the analyses of PWR core

    International Nuclear Information System (INIS)

    Tsujimoto, Iwao; Naito, Yoshitaka.

    1992-11-01

    This report is one of the materials for the work titled 'Development of the computer code system for the analyses of PWR core phenomena', which is performed under contracts between Shikoku Electric Power Company and JAERI. In this report, the numerical method adopted in our computer code system are described, that is, 'The basic course and the summary of the analysing method', 'Numerical method for solving the Boltzmann equation', 'Numerical method for solving the thermo-hydraulic equations' and 'Description on the computer code system'. (author)

  16. Degraded core accidents: review of aerosol behaviour in the containment of a PWR

    International Nuclear Information System (INIS)

    Nichols, A.L.; Walker, B.C.

    1981-09-01

    Low probability-high consequence accidents have become an important issue in reactor safety studies. Such accidents would involve damage to the core and the subsequent release of radioactive fission products into the environment. Aerosols play a major role in the transport and removal of these fission products in the reactor building containment. The aerosol mechanisms, computer modelling codes and experimental studies used to predict aerosol behaviour in the containment of a PWR are reviewed. There are significant uncertainties in the aerosol source terms and specific recommendations have been made for further studies, particularly with respect to code development and high density aerosol-fission product transport within closed systems. (author)

  17. Dynamic structural analysis for assemblies of fuel elements in the core of a PWR

    International Nuclear Information System (INIS)

    Silva Macedo, L.V. da.

    1991-01-01

    It is presented a procedure for the dynamic structural analysis of a PWR core. Impacts between fuel assemblies may occur because of the existence of gaps between them. Thus, the problem is non-linear and an spectral analysis is avoided. It is necessary a time-history response analysis. The Modal Superposition Method with the Duhamel integral was used in order to solve the problem. It is presented an algorithm of solution and also results obtained with the STYCA computer program, developed in the basis of what was proposed here. (author)

  18. Simulation of nonlinear dynamics of a PWR core by an improved lumped formulation for fuel heat transfer

    International Nuclear Information System (INIS)

    Su, Jian; Cotta, Renato M.

    2000-01-01

    In this work, thermohydraulic behaviour of PWR, during reactivity insertion and partial loss-of-flow, is simulated by using a simplified mathematical model of reactor core and primary coolant. An improved lumped parameter formulation for transient heat conduction in fuel rod is used for core heat transfer modelling. Transient temperature response of fuel, cladding and coolant is analysed. (author)

  19. NODAL3 Sensitivity Analysis for NEACRP 3D LWR Core Transient Benchmark (PWR

    Directory of Open Access Journals (Sweden)

    Surian Pinem

    2016-01-01

    Full Text Available This paper reports the results of sensitivity analysis of the multidimension, multigroup neutron diffusion NODAL3 code for the NEACRP 3D LWR core transient benchmarks (PWR. The code input parameters covered in the sensitivity analysis are the radial and axial node sizes (the number of radial node per fuel assembly and the number of axial layers, heat conduction node size in the fuel pellet and cladding, and the maximum time step. The output parameters considered in this analysis followed the above-mentioned core transient benchmarks, that is, power peak, time of power peak, power, averaged Doppler temperature, maximum fuel centerline temperature, and coolant outlet temperature at the end of simulation (5 s. The sensitivity analysis results showed that the radial node size and maximum time step give a significant effect on the transient parameters, especially the time of power peak, for the HZP and HFP conditions. The number of ring divisions for fuel pellet and cladding gives negligible effect on the transient solutions. For productive work of the PWR transient analysis, based on the present sensitivity analysis results, we recommend NODAL3 users to use 2×2 radial nodes per assembly, 1×18 axial layers per assembly, the maximum time step of 10 ms, and 9 and 1 ring divisions for fuel pellet and cladding, respectively.

  20. Optimization of refueling-shuffling scheme in PWR core by random search strategy

    International Nuclear Information System (INIS)

    Wu Yuan

    1991-11-01

    A random method for simulating optimization of refueling management in a pressurized water reactor (PWR) core is described. The main purpose of the optimization was to select the 'best' refueling arrangement scheme which would produce maximum economic benefits under certain imposed conditions. To fulfill this goal, an effective optimization strategy, two-stage random search method was developed. First, the search was made in a manner similar to the stratified sampling technique. A local optimum can be reached by comparison of the successive results. Then the other random experiences would be carried on between different strata to try to find the global optimum. In general, it can be used as a practical tool for conventional fuel management scheme. However, it can also be used in studies on optimization of Low-Leakage fuel management. Some calculations were done for a typical PWR core on a CYBER-180/830 computer. The results show that the method proposed can obtain satisfactory approach at reasonable low computational cost

  1. Evaluation of the pressure difference across the core during PWR-LOCA reflood phase

    International Nuclear Information System (INIS)

    Iguchi, Tadashi; Murao, Yoshio

    1979-03-01

    The flooding rate of the core influences largely cooling of the core during the reflood phase of a PWR-LOCA. Since the void fraction of two-phase flow in the core is important determining the flooding rate, it is essential to examine this void fraction. The void fraction in the core during the reflood phase obtained by experiment was compared with those predicted by the correlations respectively of Akagawa, Nicklin, Zuber, Yeh, Griffice, Behringer and Jhonson. Only Yeh's correlation was found to be usable for the purpose. The pressure difference of the core during the reflood phase was calculated by reflood analyzing code REFLA-1D using Yeh's correlation. Following are the results: (1) During the steady-state period after quenching of the heaters, the prediction agrees within +-15% with the experiment. (2) During the transient period when the quench front is advancing, the prediction is not in agreement with the experiment, the difference being about +-40%. Influence of the advancing quench front upon the void fraction in the core must further be studied. (author)

  2. Inference of core barrel motion from neutron noise spectral density. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, J.C.; Shahrokhi, F.; Kryter, R.C.

    1977-03-15

    A method was developed for inference of core barrel motion from the following statistical descriptors: cross-power spectral density, autopower spectral density, and amplitude probability density. To quantify the core barrel motion in a typical pressurized water reactor (PWR), a scale factor was calculated in both one- and two-dimensional geometries using forward, variational, and perturbation methods of discrete ordinates neutron transport. A procedure for selection of the proper frequency band limits for the statistical descriptors was developed. It was found that although perturbation theory is adequate for the calculation of the scale factor, two-dimensional geometric effects are important enough to rule out the use of a one-dimensional approximation for all but the crudest calculations. It was also found that contributions of gamma rays can be ignored and that the results are relatively insensitive to the cross-section set employed. The proper frequency band for the statistical descriptors is conveniently determined from the coherence and phase information from two ex-core power range neutron monitors positioned diametrically across the reactor vessel. Core barrel motion can then be quantified from the integral of the band-limited cross-power spectral density of two diametrically opposed ex-core monitors or, if the coherence between the pair is greater than or equal to 0.7, from a properly band-limited amplitude probability density function. Wide-band amplitude probability density functions were demonstrated to yield erroneous estimates for the magnitude of core barrel motion.

  3. Assessment of the insertion of reprocessed fuel spiked with thorium in a PWR core

    Energy Technology Data Exchange (ETDEWEB)

    Castro, Victor F.; Monteiro, Fabiana B.A.; Pereira, Claubia, E-mail: victorfc@fis.grad.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-07-01

    Reprocessed fuel by UREX+ technique and spiked with thorium was inserted in a PWR core and neutronic parameters have been analyzed. Based on the Final Safety Analysis Report (FSAR) of the Angra-2 reactor, the core was modeled and simulated with SCALE6.0 package. The neutronic data evaluation was carried out by the analysis of the effective and infinite multiplication factors, and the fuel evolution during the burnup. The conversion ratio (CR) was also evaluated. The results show that, when inserting reprocessed fuel spiked with thorium, the insertion of burnable poison rods is not necessary, due to the amount of absorber isotopes present in the fuel. Besides, the conversion ratio obtained was greater than the presented by standard UO{sub 2} fuel, indicating the possibility of extending the burnup. (author)

  4. Research on 3D power distribution of PWR reactor core based on RBF neural network

    International Nuclear Information System (INIS)

    Xia Hong; Li Bin; Liu Jianxin

    2014-01-01

    Real-time monitor for 3D power distribution is critical to nuclear safety and high efficiency of NPP's operation as well as the control system optimization. A method was proposed to set up a real-time monitor system for 3D power distribution by using of ex-core neutron detecting system and RBF neural network for improving the instantaneity of the monitoring results and reducing the fitting error of the 3D power distribution. A series of experiments were operated on a 300 MW PWR simulation system. The results demonstrate that the new monitor system works very well under condition of certain burnup range during the fuel cycle and reconstructs the real-time 3D distribution of reactor core power. The accuracy of the model is improved effectively with the help of several methods. (authors)

  5. A novel optimization method, Gravitational Search Algorithm (GSA), for PWR core optimization

    International Nuclear Information System (INIS)

    Mahmoudi, S.M.; Aghaie, M.; Bahonar, M.; Poursalehi, N.

    2016-01-01

    Highlights: • The Gravitational Search Algorithm (GSA) is introduced. • The advantage of GSA is verified in Shekel’s Foxholes. • Reload optimizing in WWER-1000 and WWER-440 cases are performed. • Maximizing K eff , minimizing PPFs and flattening power density is considered. - Abstract: In-core fuel management optimization (ICFMO) is one of the most challenging concepts of nuclear engineering. In recent decades several meta-heuristic algorithms or computational intelligence methods have been expanded to optimize reactor core loading pattern. This paper presents a new method of using Gravitational Search Algorithm (GSA) for in-core fuel management optimization. The GSA is constructed based on the law of gravity and the notion of mass interactions. It uses the theory of Newtonian physics and searcher agents are the collection of masses. In this work, at the first step, GSA method is compared with other meta-heuristic algorithms on Shekel’s Foxholes problem. In the second step for finding the best core, the GSA algorithm has been performed for three PWR test cases including WWER-1000 and WWER-440 reactors. In these cases, Multi objective optimizations with the following goals are considered, increment of multiplication factor (K eff ), decrement of power peaking factor (PPF) and power density flattening. It is notable that for neutronic calculation, PARCS (Purdue Advanced Reactor Core Simulator) code is used. The results demonstrate that GSA algorithm have promising performance and could be proposed for other optimization problems of nuclear engineering field.

  6. Core losses of an inverter-fed permanent magnet synchronous motor with an amorphous stator core under no-load

    Directory of Open Access Journals (Sweden)

    Nicolas Denis

    2016-05-01

    Full Text Available In this paper, an interior permanent magnet synchronous motor (IPMSM with a stator core made of amorphous magnetic material (AMM is presented. The IPMSM is driven by a voltage source three-phase inverter with classical pulse width modulation (PWM control. The core losses under no-load condition are measured by experiment and compared to an equivalent IPMSM with a stator core made of NO steel. Under these conditions, the core losses are influenced by the stator, rotor and magnet shapes but also by the PWM carrier signal that implies a high frequency harmonic in the magnetic flux density. It is demonstrated that the AMM can reduce the core losses by about 56 %.

  7. Reflooding phenomena of German PWR estimated from CCTF [Cylindrical Core Test Facility], SCTF [Slab Core Test Facility] and UPTF [Upper Plenum Test Facility] results

    International Nuclear Information System (INIS)

    Murao, Y.; Iguchi, T.; Sugimoto, J.

    1988-09-01

    The reflooding behavior in a PWR with a combined injection type ECCS was studied by comparing the test results from Cylindrical Core Test Facility (CCTF), Slab Core Test Facility (SCTF) and Upper Plenum Test Facility (UPTF). Core thermal-hydraulics is discussed mainly based on SCTF test data. In addition, the water accumulation behavior in hot legs and the break-through characteristics at tie plate are discussed

  8. Determination of PWR core water level using ex-core detectors signals

    International Nuclear Information System (INIS)

    Bernal, Alvaro; Abarca, Agustin; Miro, Rafael; Verdu, Gumersindo

    2013-01-01

    The core water level provides relevant neutronic and thermalhydraulic information of the reactor such as power, k eff and cooling ability; in fact, core water level monitoring could be used to predict LOCA and cooling reduction which may deal with core damage. Although different detection equipment is used to monitor several parameters such as the power, core water level monitoring is not an evident task. However, ex-core detectors can measure the fast neutrons leaking the core and several studies demonstrate the existence of a relationship between fast neutron leakage and core water level due to the shielding effect of the water. In addition, new ex-core detectors are being developed, such as silicon carbide semiconductor radiation detectors, monitoring the neutron flux with higher accuracy and in higher temperatures conditions. Therefore, a methodology to determine this relationship has been developed based on a Monte Carlo calculation using MCNP code and applying variance reduction with adjoint functions based on the adjoint flux obtained with the discrete ordinates code TORT. (author)

  9. 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)

  10. Turbulent heat transfer in a coolant channel of a pressurized water reactor (PWR) core

    International Nuclear Information System (INIS)

    Kumar, Sanjeev; Saha, Arun K.; Munshi, Prabhat

    2016-01-01

    Exact predictions in nuclear reactors are more crucial, because of the safety aspects. It necessitates the appropriate modeling of heat transfer phenomena in the reactors core. A two-dimensional thermal-hydraulics model is used to study the detailed analysis of the coolant region of a fuel pin. Governing equations are solved using Marker and Cell (MAC) method. Standard wall functions k-ε turbulence model is incorporated to consider the turbulent behaviour of the flow field. Validation of the code and a few results for a typical PWR running at normal operating conditions reported earlier. There were some discrepancies in the old calculations. These discrepancies have been resolved and updated results are presented in this work. 2D thermal-hydraulics model results have been compared with the 1D thermal-hydraulics model results and conclusions have been drawn. (author)

  11. Experimental modelling of core debris dispersion from the vault under a PWR pressure vessel. Pt. 2

    International Nuclear Information System (INIS)

    Rose, P.W.

    1987-12-01

    In previous experiments, done on a 1/25 scale model in Perspex of the vault under a PWR pressure vessel, the instrument tubes support structure built into the vault was not included. It consists of a number of grids made up of fairly massive steel girders. These have now been added to the model and experiments performed using water to simulate molten core debris assumed to have fallen on to the vault floor and high-pressure air to simulate the discharge of steam or gas from the assumed breach at the bottom of the pressure vessel. The results show that the tubes support structure considerably reduces the carry-over of liquid via the vault access shafts. (author)

  12. Degraded core accidents for the Sizewell PWR A sensitivity analysis of the radiological consequences

    CERN Document Server

    Kelly, G N; Clarke, R H; Ferguson, L; Haywood, S M; Hemming, C R; Jones, J A

    1982-01-01

    The radiological impact of degraded core accidents postulated for the Sizewell PWR was assessed in an earlier study. In this report the sensitivity of the predicted consequences to variation in the values of a number of important parameters is investigated for one of the postulated accidental releases. The parameters subjected to sensitivity analyses are the dose-mortality relationship for bone marrow irradiation, the energy content of the release, the warning time before the release to the environment, and the dry deposition velocity for airborne material. These parameters were identified as among the more important in determining the uncertainty in the results obtained in the initial study. With a few exceptions the predicted consequences were found to be not very sensitive to the parameter values investigated, the range of variation in the consequences for the limiting values of each parameter rarely exceeded a factor of a few and in many cases was considerably less. The conclusions reached are, however, p...

  13. On-line thermal margin estimation of a PWR core using a neural network approach

    International Nuclear Information System (INIS)

    Park, Soon Ok; Kim, Hyun Koon; Lee, Seung Hynk; Chang, Soon Heung

    1992-01-01

    A new approach for on-line thermal margin monitoring of a PWR Core is proposed in this paper, where a neural network model is introduced to predict the DNBR values at the given reactor operating conditions. The neural network is learned by the Back Propagation algorithm with the optimized random training data and is tested to investigate the generalized performance for the steady state operating region as well as for the transient situations where DNB is of the primary concern. The test results show that the high level of accuracy in predicting the DNBR can be achieved by the neural network model compared to the detailed code results. An insight has been gained from this study that the neural network model for estimating DNB performance can be a viable tool for on-line thermal margin monitoring of a nuclear power plant

  14. 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

  15. An axial calculation method for accurate two-dimensional PWR core simulation

    International Nuclear Information System (INIS)

    Grimm, P.

    1985-02-01

    An axial calculation method, which improves the agreement of the multiplication factors determined by two- and three-dimensional PWR neutronic calculations, is presented. The axial buckling is determined at each time point so as to reproduce the increase of the leakage due to the flattening of the axial power distribution and the effect of the axial variation of the group constants of the fuel on the reactivity is taken into account. The results of a test example show that the differences of k-eff and cycle length between two- and three-dimensional calculations, which are unsatisfactorily large if a constant buckling is used, become negligible if the results of the axial calculation are used in the two-dimensional core simulation. (Auth.)

  16. 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

  17. Global calculation of PWR reactor core using the two group energy solution by the response matrix method

    International Nuclear Information System (INIS)

    Conti, C.F.S.; Watson, F.V.

    1991-01-01

    A computational code to solve a two energy group neutron diffusion problem has been developed base d on the Response Matrix Method. That method solves the global problem of PWR core, without using the cross sections homogenization process, thus it is equivalent to a pontwise core calculation. The present version of the code calculates the response matrices by the first order perturbative method and considers developments on arbitrary order Fourier series for the boundary fluxes and interior fluxes. (author)

  18. Mixed PWR core loadings with inert matrix Pu-fuel assemblies

    International Nuclear Information System (INIS)

    Stanculescu, A.; Kasemeyer, U.; Paratte, J.-M.; Chawla, R.

    1999-01-01

    The most efficient way to enhance plutonium consumption in light water reactors is to eliminate the production of plutonium all together. This requirement leads to fuel concepts in which the uranium is replaced by an inert matrix. At PSI, studies have focused on employing ZrO 2 as inert matrix. Adding a burnable poison to such a fuel proves to be necessary. As a result of scoping studies, Er 2 O 3 was identified as the most suitable burnable poison material. The results of whole-core three-dimensional neutronics analyses indicated, for a present-day 1000 MW e pressurised water reactor, the feasibility of an asymptotic equilibrium four-batch cycle fuelled solely with the proposed PuO 2 -Er 2 O 3 -ZrO 2 inert matrix fuel (IMF). The present paper presents the results of more recent investigations related to 'real-life' situations, which call for transition configurations in which mixed IMF and UO 2 assembly loadings must be considered. To determine the influence of the introduction of IMF assemblies on the characteristics of a UO 2 -fuelled core, three-dimensional full-core calculations have been performed for a present-day 1000 MW e PWR containing up to 12 optimised IMF assemblies. (author)

  19. 3D thermal-hydraulic analysis on core of PWR nuclear power station

    International Nuclear Information System (INIS)

    Yao Zhaohui; Wang Xuefang; Shen Mengyu

    1997-01-01

    Thermal hydraulic analysis of core is of great importance in reactor safety analysis. A computer code, thermal hydraulic analysis porous medium analysis (THAPMA), has been developed to simulate the flow and heat transfer characteristics of reactor components. It has been proved reliable by several numerical tests. In the THAPMA code, a new difference scheme and solution method have been studied in developing the computer software. For the difference scheme, a second order accurate, high resolution scheme, called WSUC scheme, has been proposed. This scheme is total variation bounded and unconditionally stable in convective numeral stability. Numerical tests show that the WSUC is better in accuracy and resolution than the 1-st order upwind, 2-nd order upwind, SOUCUP by Zhu and Rodi. In solution method, a modified PISO algorithm is used, which is not only simpler but also more accurate and more rapid in convergence than the original PISO algorithm. Moreover, the modified PISO algorithm can effectively solve steady and transient state problem. Besides, with the THAPMA code, the flow and heat transfer phenomena in reactor core have been numerically simulated in the light of the design condition of Qinshan PWR nuclear power station (the second-term project). The simulation results supply a theoretical basis for the core design

  20. PWR core and spent fuel pool analysis using scale and nestle

    International Nuclear Information System (INIS)

    Murphy, J. E.; Maldonado, G. I.; St Clair, R.; Orr, D.

    2012-01-01

    The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

  1. PWR core and spent fuel pool analysis using scale and nestle

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, J. E.; Maldonado, G. I. [Dept. of Nuclear Engineering, Univ. of Tennessee, Knoxville, TN 37996-2300 (United States); St Clair, R.; Orr, D. [Duke Energy, 526 S. Church St, Charlotte, NC 28202 (United States)

    2012-07-01

    The SCALE nuclear analysis code system [SCALE, 2011], developed and maintained at Oak Ridge National Laboratory (ORNL) is widely recognized as high quality software for analyzing nuclear systems. The SCALE code system is composed of several validated computer codes and methods with standard control sequences, such as the TRITON/NEWT lattice physics sequence, which supplies dependable and accurate analyses for industry, regulators, and academia. Although TRITON generates energy-collapsed and space-homogenized few group cross sections, SCALE does not include a full-core nodal neutron diffusion simulation module within. However, in the past few years, the open-source NESTLE core simulator [NESTLE, 2003], originally developed at North Carolina State Univ. (NCSU), has been updated and upgraded via collaboration between ORNL and the Univ. of Tennessee (UT), so it now has a growingly seamless coupling to the TRITON/NEWT lattice physics [Galloway, 2010]. This study presents the methodology used to couple lattice physics data between TRITON and NESTLE in order to perform a three-dimensional full-core analysis employing a 'real-life' Duke Energy PWR as the test bed. The focus for this step was to compare the key parameters of core reactivity and radial power distribution versus plant data. Following the core analysis, following a three cycle burn, a spent fuel pool analysis was done using information generated from NESTLE for the discharged bundles and was compared to Duke Energy spent fuel pool models. The KENO control module from SCALE was employed for this latter stage of the project. (authors)

  2. Integral Full Core Multi-Physics PWR Benchmark with Measured Data

    Energy Technology Data Exchange (ETDEWEB)

    Forget, Benoit; Smith, Kord; Kumar, Shikhar; Rathbun, Miriam; Liang, Jingang

    2018-04-11

    In recent years, the importance of modeling and simulation has been highlighted extensively in the DOE research portfolio with concrete examples in nuclear engineering with the CASL and NEAMS programs. These research efforts and similar efforts worldwide aim at the development of high-fidelity multi-physics analysis tools for the simulation of current and next-generation nuclear power reactors. Like all analysis tools, verification and validation is essential to guarantee proper functioning of the software and methods employed. The current approach relies mainly on the validation of single physic phenomena (e.g. critical experiment, flow loops, etc.) and there is a lack of relevant multiphysics benchmark measurements that are necessary to validate high-fidelity methods being developed today. This work introduces a new multi-cycle full-core Pressurized Water Reactor (PWR) depletion benchmark based on two operational cycles of a commercial nuclear power plant that provides a detailed description of fuel assemblies, burnable absorbers, in-core fission detectors, core loading and re-loading patterns. This benchmark enables analysts to develop extremely detailed reactor core models that can be used for testing and validation of coupled neutron transport, thermal-hydraulics, and fuel isotopic depletion. The benchmark also provides measured reactor data for Hot Zero Power (HZP) physics tests, boron letdown curves, and three-dimensional in-core flux maps from 58 instrumented assemblies. The benchmark description is now available online and has been used by many groups. However, much work remains to be done on the quantification of uncertainties and modeling sensitivities. This work aims to address these deficiencies and make this benchmark a true non-proprietary international benchmark for the validation of high-fidelity tools. This report details the BEAVRS uncertainty quantification for the first two cycle of operations and serves as the final report of the project.

  3. PWR core design, neutronics evaluation and fuel cycle analysis for thorium-uranium breeding recycle

    International Nuclear Information System (INIS)

    Bi, G.; Liu, C.; Si, S.

    2012-01-01

    This paper was focused on core design, neutronics evaluation and fuel cycle analysis for Thorium-Uranium Breeding Recycle in current PWRs, without any major change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. The fuel cycle analysis indicates that Thorium-Uranium Breeding Recycle is technically feasible in current PWRs. A 4-loop, 193-assembly PWR core utilizing 17 x 17 fuel assemblies (FAs) was taken as the model core. Two mixed cores were investigated respectively loaded with mixed reactor grade Plutonium-Thorium (PuThOX) FAs and mixed reactor grade 233 U-Thorium (U 3 ThOX) FAs on the basis of reference full Uranium oxide (UOX) equilibrium-cycle core. The UOX/PuThOX mixed core consists of 121 UOX FAs and 72 PuThOX FAs. The reactor grade 233 U extracted from burnt PuThOX fuel was used to fabrication of U 3 ThOX for starting Thorium-. Uranium breeding recycle. In UOX/U 3 ThOX mixed core, the well designed U 3 ThOX FAs with 1.94 w/o fissile uranium (mainly 233 U) were located on the periphery of core as a blanket region. U 3 ThOX FAs remained in-core for 6 cycles with the discharged burnup achieving 28 GWD/tHM. Compared with initially loading, the fissile material inventory in U 3 ThOX fuel has increased by 7% via 1-year cooling after discharge. 157 UOX fuel assemblies were located in the inner of UOX/U 3 ThOX mixed core refueling with 64 FAs at each cycle. The designed UOX/PuThOX and UOX/U 3 ThOX mixed core satisfied related nuclear design criteria. The full core performance analyses have shown that mixed core with PuThOX loading has similar impacts as MOX on several neutronic characteristic parameters, such as reduced differential boron worth, higher critical boron concentration, more negative moderator temperature coefficient, reduced control rod worth, reduced shutdown margin, etc.; while mixed core with U 3 ThOX loading on the periphery of core has no visible impacts on neutronic characteristics compared

  4. Method of injecting cooling water in emergency core cooling system (ECCS) of PWR type reactor

    International Nuclear Information System (INIS)

    Sobajima, Makoto; Adachi, Michihiro; Tasaka, Kanji; Suzuki, Mitsuhiro.

    1979-01-01

    Purpose: To provide a cooling water injection method in an ECCS, which can perform effective cooling of the reactor core. Method: In a method of injecting cooling water in an ECCS as a countermeasure against a rupture accident of a pwr type reactor, cooling water in the first pressure storage injection system is injected into the upper plenum of the reactor pressure vessel at a set pressure of from 50 to 90 atg. and a set temperature of from 80 to 200 0 C, cooling water in the second pressure storage injection system is injected into the lower plenum of the reactor pressure vessel at a pressure of from 25 to 60 atg. which is lower than the set pressure and a temperature less than 60 0 C, and further in combination with these procedures, cooling water of less than 60 0 C is injected into a high-temperature side piping, in the high-pressure injection system of upstroke of 100 atg. by means of a pump and the low-pressure injection system of upstroke of 20 atg. also by means of a pump, thereby cooling the reactor core. (Aizawa, K.)

  5. Study of corium radial spreading between fuel rods in a PWR core

    International Nuclear Information System (INIS)

    Roche, S.; Gatt, J.M.

    1996-01-01

    In the framework of severe accident studies for PWR like Three Mile Island Unit 2 (TMI-2), the reactor core essentially constituted of fuel rods begins to heat and then to melt. During the early degradation phase, a melt (essentially UO2 and ZrO2) that constitutes the corium flows first along the rods, and after a blockage formation, may radially propagate towards the core periphery. A simplified model has been elaborated to study the corium freezing phenomena during its crossflow between the fuel rods. The corium spreads on an horizontal support made, of either a corium crust, or a grid assembly. The model solves numerically the interface energy balance equation at the solid-liquid corium interface and the monodimensional heat balance equation in transient process with convective terms and heat source (residual power). ''Zukauskas'' correlations are used to calculate heat transfer coefficients. The model can be integrated in severe accident codes like ICARE II (IPSN) describing the in-vessel degradation scenarios. (author). 5 refs, 10 figs

  6. New long-cycle small modular PWR cores using particle type burnable poisons for low boron operation

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Hoseong; Hwang, Dae Hee [Department of Nuclear Engineering, Kyung Hee University, Deogyeong-daero, GiHeung-gu, Yongin, Gyeonggi-do 446-701 (Korea, Republic of); Hong, Ser Gi, E-mail: sergihong@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, Deogyeong-daero, GiHeung-gu, Yongin, Gyeonggi-do 446-701 (Korea, Republic of); Shin, Ho Choel [Core and Fuel Analysis Group, Korea Hydro & Nuclear Power Central Research Institute (KHNP-CRI), Daejon 305-343 (Korea, Republic of)

    2017-04-01

    Highlights: • New advanced burnable poison rods (BPR) are suggested for low boron operation in PWR. • The new SMR cores have long cycle length of ∼4.5 EFPYs with low boron concentration. • The SMR core satisfies all the design targets and constraints. - Abstract: In this paper, new small long-cycle PWR (Pressurized Water Reactor) cores for low boron concentration operation are designed by employing advanced burnable poison rods (BPRs) in which the BISO (Bi-Isotropic) particles of burnable poison are distributed in a SiC matrix. The BPRs are designed by adjusting the kernel diameter, the kernel material and the packing fraction to effectively reduce the excess reactivity in order to reduce the boron concentration in the coolant and achieve a flat change in excess reactivity over a long operational cycle. In addition, axial zoning of the BPRs was suggested to improve the core performances, and it was shown that the suggested axial zoning of BPRs considerably extends the cycle length compared to a core with no BPR axial zoning. The results of the core physics analyses showed that the cores using BPRs with a B{sub 4}C kernel have long cycle lengths of ∼4.5 EFPYs (Effective Full Power Years), small maximum CBCs (Critical Boron Concentration) lower than 370 ppm, low power peaking factors, and large shutdown margins of control element assemblies.

  7. Layout of PWR in-core instrumentation system tubing and support structure with Bechtel 3D-CADD

    International Nuclear Information System (INIS)

    Ichikawa, T.; Pfeifer, B.W.; Mulay, J.N.

    1987-01-01

    The optimization study of the PWR In-Core Instrumentation System (ICIS) tubing layout and support structure presented an opportunity to utilize the Bechtel 3D-CADD program to perform this task. This paper provides a brief summary of the Bechtel 3D-CADD program development and capabilities and outlines the process of developing and optimizing the ICIS tube layout. Specific aspects relating to the ICIS tube layout criteria, support, alignment, electronic interference check and erection sequence are provided. (orig.)

  8. Assessment of fission product release from the reactor containment building during severe core damage accidents in a PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.; Generino, G.

    1984-07-01

    Fission product releases from the RCB associated with hypothetical core-melt accidents ABβ, S 2 CDβ and TLBβ in a PWR-900 MWe have been performed using French computer codes (in particular, the JERICHO Code for containment response analysis and AEROSOLS/B1 for aerosol behavior in the containment) related to thermalhydraulics and fission product behavior in the primary system and in the reactor containment building

  9. Analysis of Moderator Temperature Reactivity Coefficient of the PWR Core Using WIMS-ANL

    International Nuclear Information System (INIS)

    Tukiran; Rokhmadi

    2007-01-01

    The Moderator Temperature Reactivity Coefficient (MTRC) is an important parameter in design, control and safety, particularly in PWR reactor. It is then very important to validate any new processed library for an accurate prediction of this parameter. The objective of this work is to validate the newly WIMS library based on ENDF/B-VI nuclear data files, especially for the prediction of the MTRC parameter. For this purpose, it is used a set of light water moderated lattice experiments as the NORA experiment and R1-100H critical reactors, both of reactors using UO 2 fuel pellet. Analysis is used with WIMSD/4 lattice code with original cross section libraries and WIMS-ANL with ENDF/B-VI cross section libraries. The results showed that the moderator temperatures reactivity coefficients for the NORA reactor using original libraries is - 5.039E-04 %Δk/k/℃ but for ENDF/B-VI libraries is - 2.925E-03 %Δk/k/℃. Compared to the designed value of the reactor core, the difference is in the range of 1.8 - 3.8 % for ENDF/B-IV libraries. It can be concluded that for reactor safety and control analysis, it has to be used ENDF/B- VI libraries because the original libraries is not accurate any more. (author)

  10. Analysis of a small PWR core with the PARCS/Helios and PARCS/Serpent code systems

    International Nuclear Information System (INIS)

    Baiocco, G.; Petruzzi, A.; Bznuni, S.; Kozlowski, T.

    2017-01-01

    Highlights: • The consistency between Helios and Serpent few-group cross sections is shown. • The PARCS model is validated against a Monte Carlo 3D model. • The fission and capture rates are compared. • The influence of the spacer grids on the axial power distribution is shown. - Abstract: Lattice physics codes are primarily used to generate cross-section data for nodal codes. In this work the methodology of homogenized constant generation was applied to a small Pressurized Water Reactor (PWR) core, using the deterministic code Helios and the Monte Carlo code Serpent. Subsequently, a 3D analysis of the PWR core was performed with the nodal diffusion code PARCS using the two-group cross section data sets generated by Helios and Serpent. Moreover, a full 3D model of the PWR core was developed using Serpent in order to obtain a reference solution. Several parameters, such as k eff , axial and radial power, fission and capture rates were compared and found to be in good agreement.

  11. Analysis of the core reflooding of a PWR reactor under a loss-of-coolant postulated accident

    International Nuclear Information System (INIS)

    Austregesilo Filho, H.

    1978-12-01

    The main purpose of this work is to analyse the termohydraulic behaviour of emergency cooling water, during reflooding of a PWR core submitted to a postulated loss-of-coolant accident, with the scope of giving the boundary conditions needed to verify fuel element and containment integrity. The analytical model presented was applied to the simulation of Angra I core reflooding phase, after a double-ended break between pressure vessel and discharge of one of the main coolant pumps. For this accident, with a discharge coefficient of C sub(D) = 0.4, the highest peak cladding temperature is expected. (author) [pt

  12. 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.

  13. 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.

  14. Impact forces on a core shroud of an excited PWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Collard, B.; Vallory, J. [CEA Cadarache, 13 - Saint Paul lez Durance (France)

    2001-07-01

    Seismic excitation of PWR internals may induce large motions of the fuel assemblies (FA). This could result in impact between assemblies or between assemblies and core shroud. Forces generated during these shocks are often the basis for the maximum design loads of the spacer grids and fuel rods. An experimental program has been conducted at the French Nuclear Reactor Directorate (CEA) to measure the impact forces of a reduced scale FA on the test section under different environmental conditions. Within the framework of the tests presented, the effect of the FA environment (air, stagnant water, water under flow) on the maximum impact forces measured at grid levels and on the energy dissipated during the shock is examined. A 'fluid cushioning' effect (dissipative) between the grids and the wall is sought. Experimental results show that the axial flow has a great influence on the impact forces. The greater the axial flow velocity is, the lower the impact forces are. The tests of impact of an assembly on a wall were analyzed compared to the tests carried out without impact. This analysis related on the measured forces of impact and the variation of the measured/computed total energy of the system. The whole of these tests in air and water shows that the 'fluid cushioning' effect required exists but is not significant. Thus the presence of water does not decrease the forces of impact, and does not amplify the quantity of energy dissipated during the shock. The fact that the 'fluid cushioning' effect is weak compared to more analytical tests probably comes from our 'not perfect' or 'realistic' conditions of tests which involve an angle between the grid and the wall at the shock moment.

  15. Impact forces on a core shroud of an excited PWR fuel assembly

    International Nuclear Information System (INIS)

    Collard, B.; Vallory, J.

    2001-01-01

    Seismic excitation of PWR internals may induce large motions of the fuel assemblies (FA). This could result in impact between assemblies or between assemblies and core shroud. Forces generated during these shocks are often the basis for the maximum design loads of the spacer grids and fuel rods. An experimental program has been conducted at the French Nuclear Reactor Directorate (CEA) to measure the impact forces of a reduced scale FA on the test section under different environmental conditions. Within the framework of the tests presented, the effect of the FA environment (air, stagnant water, water under flow) on the maximum impact forces measured at grid levels and on the energy dissipated during the shock is examined. A 'fluid cushioning' effect (dissipative) between the grids and the wall is sought. Experimental results show that the axial flow has a great influence on the impact forces. The greater the axial flow velocity is, the lower the impact forces are. The tests of impact of an assembly on a wall were analyzed compared to the tests carried out without impact. This analysis related on the measured forces of impact and the variation of the measured/computed total energy of the system. The whole of these tests in air and water shows that the 'fluid cushioning' effect required exists but is not significant. Thus the presence of water does not decrease the forces of impact, and does not amplify the quantity of energy dissipated during the shock. The fact that the 'fluid cushioning' effect is weak compared to more analytical tests probably comes from our 'not perfect' or 'realistic' conditions of tests which involve an angle between the grid and the wall at the shock moment

  16. Synthesis and magnetic properties of inverted core-shell polyaniline-ferrite composite

    Energy Technology Data Exchange (ETDEWEB)

    Donescu, Dan [National Institute for Research & Development in Chemistry and Petrochemistry - ICECHIM, 202 Spl. Independentei, 060021, Bucharest (Romania); Fierascu, Radu Claudiu, E-mail: radu_claudiu_fierascu@yahoo.com [National Institute for Research & Development in Chemistry and Petrochemistry - ICECHIM, 202 Spl. Independentei, 060021, Bucharest (Romania); Ghiurea, Marius [National Institute for Research & Development in Chemistry and Petrochemistry - ICECHIM, 202 Spl. Independentei, 060021, Bucharest (Romania); Manaila-Maximean, Doina [University Politehnica of Bucharest, Department of Physics, 313 Spl. Independentei, 060042, Bucharest (Romania); Nicolae, Cristian Andi; Somoghi, Raluca; Spataru, Catalin Ilie [National Institute for Research & Development in Chemistry and Petrochemistry - ICECHIM, 202 Spl. Independentei, 060021, Bucharest (Romania); Stanica, Nicolae [Institute of Physical Chemistry “Ilie Murgulescu”, 202 Spl. Independentei, 060021, Bucharest (Romania); Raditoiu, Valentin [National Institute for Research & Development in Chemistry and Petrochemistry - ICECHIM, 202 Spl. Independentei, 060021, Bucharest (Romania); Vasile, Eugeniu [SC METAV – CD SA, 31 C. A. Rosetti Str., 021051, Bucharest (Romania)

    2017-08-31

    The present paper studies the effect of polyaniline grafting on magnetite functionalized with aminopropyltrimethoxysilane. All the compounds were characterized by analytical techniques (X-ray fluorescence, X-ray diffraction, Fourier transform infrared spectroscopy, thermal analysis, Transmission electron microscopy), as well as by determining their magnetic properties. The electron microscopy analysis of the hybrids shows similar morphologies for all the samples. The presence of the iron atoms on the surface of the final product supports the idea of the existence of an inverted core-shell type structure, the more polar ferrite orienting itself towards water. The correlation between the maximum grafting probability and the maximum magnetization is evidenced, demonstrating the importance of the polymer grafting method on the magnetic properties.

  17. The radiological consequences of degraded core accidents for the Sizewell PWR The impact of adopting revised frequencies of occurrence

    CERN Document Server

    Kelly, G N

    1983-01-01

    The radiological consequences of degraded core accidents postulated for the Sizewell PWR were assessed in an earlier study and the results published in NRPB-R137. Further analyses have since been made by the Central Electricity Generating Board (CEGB) of degraded core accidents which have led to a revision of their predicted frequencies of occurrence. The implications of these revised frequencies, in terms of the risk to the public from degraded core accidents, are evaluated in this report. Increases, by factors typically within the range of about 1.5 to 7, are predicted in the consequences, compared with those estimated in the earlier study. However, the predicted risk from degraded core accidents, despite these increases, remains exceedingly small.

  18. Three-dimensional transport coefficient model and prediction-correction numerical method for thermal margin analysis of PWR cores

    International Nuclear Information System (INIS)

    Chiu, C.

    1981-01-01

    Combustion Engineering Inc. designs its modern PWR reactor cores using open-core thermal-hydraulic methods where the mass, momentum and energy equations are solved in three dimensions (one axial and two lateral directions). The resultant fluid properties are used to compute the minimum Departure from Nuclear Boiling Ratio (DNBR) which ultimately sets the power capability of the core. The on-line digital monitoring and protection systems require a small fast-running algorithm of the design code. This paper presents two techniques used in the development of the on-line DNB algorithm. First, a three-dimensional transport coefficient model is introduced to radially group the flow subchannel into channels for the thermal-hydraulic fluid properties calculation. Conservation equations of mass, momentum and energy for this channels are derived using transport coefficients to modify the calculation of the radial transport of enthalpy and momentum. Second, a simplified, non-iterative numerical method, called the prediction-correction method, is applied together with the transport coefficient model to reduce the computer execution time in the determination of fluid properties. Comparison of the algorithm and the design thermal-hydraulic code shows agreement to within 0.65% equivalent power at a 95/95 confidence/probability level for all normal operating conditions of the PWR core. This algorithm accuracy is achieved with 1/800th of the computer processing time of its parent design code. (orig.)

  19. Core radial power profile effect on system and core cooling behavior during reflood phase of PWR-LOCA with CCTF data

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Iguchi, Tadashi; Murao, Yoshio

    1985-01-01

    In the reactor safety assessment during reflood phase of a PWR-LOCA, it is assumed implicitly that the core thermal hydraulic behavior is evaluated by the one-dimensional model with an average power rod. In order to assess the applicability of the one-dimensional treatment, integral tests were performed with various core radial power profiles using the Cylindrical Core Test Facility (CCTF) whose core includes about 2,000 heater rods. The CCTF results confirm that the core radial power profile has weak effect on the thermal hydraulic behavior in the primary system except core. It is also confirmed that the core differential pressure in the axial direction is predicted by the one-dimensional core model with an average power rod even in the case with a steep radial power profile in the core. Even though the core heat transfer coefficient is dependent on the core radial power profile, it is found that the error of the peak clad surface temperature calculation is less than 15 K using the one-dimensional model in the CCTF tests. The CCTF results support the one-dimensional treatment assumed in the reactor safety assessment. (author)

  20. The coupling of the Star-Cd software to a whole-core neutron transport code Decart for PWR applications

    International Nuclear Information System (INIS)

    Thomas, J.W.; Lee, H.C.; Downar, T.J.; Sofu, T.; Weber, D.P.; Joo, H.G.; Cho, J.Y.

    2003-01-01

    As part of a U.S.- Korea collaborative U.S. Department of Energy INERI project, a comprehensive high-fidelity reactor-core modeling capability is being developed for detailed analysis of existing and advanced PWR reactor designs. An essential element of the project has been the development of an interface between the computational fluid dynamics (CFD) module, STAR-CD, and the neutronics module, DeCART. Since the computational mesh for CFD and neutronics calculations are generally different, the capability to average and decompose data on these different meshes has been an important part of code coupling activities. An averaging process has been developed to extract neutronics zone temperatures in the fuel and coolant and to generate appropriate multi group cross sections and densities. Similar procedures have also been established to map the power distribution from the neutronics zones to the mesh structure used in the CFD module. Since MPI is used as the parallel model in STAR-CD and conflicts arise during initiation of a second level of MPI, the interface developed here is based on using TCP/IP protocol sockets to establish communication between the CFD and neutronics modules. Preliminary coupled calculations have been performed for PWR fuel assembly size problems and converged solutions have been achieved for a series of steady-state problems ranging from a single pin to a 1/8 model of a 17 x 17 PWR fuel assembly. (authors)

  1. ASCOT-1: a computer program for analyzing the thermo-hydraulic behavior in a PWR core during a LOCA

    International Nuclear Information System (INIS)

    Kobayashi, Kensuke; Sato, Kazuo

    1978-09-01

    A digital computer code ASCOT-1 has been developed to analyze the thermo-hydraulic behavior in a PWR core during a loss-of-coolant accident. The core is assumed to be axi-symmetric two-dimensional and the conservation laws are solved by the method of characteristics. For the temperature response of representative fuels of the concentric annular subregions into which the core is divided, the heat conduction equations are solved by the explicit method with the averaged flow conditions decided above. The boundary conditions at the upper and lower plenum are given as inputs. The program is of an adjustable dimension so there are no restrictions to the numbers of meshes. ASCOT-1 is written in FORTRAN-IV for FACOM230-75. (author)

  2. VALIDATION OF FULL CORE GEOMETRY MODEL OF THE NODAL3 CODE IN THE PWR TRANSIENT BENCHMARK PROBLEMS

    Directory of Open Access Journals (Sweden)

    Tagor Malem Sembiring

    2015-10-01

    Full Text Available ABSTRACT VALIDATION OF FULL CORE GEOMETRY MODEL OF THE NODAL3 CODE IN THE PWR TRANSIENT BENCHMARK PROBLEMS. The coupled neutronic and thermal-hydraulic (T/H code, NODAL3 code, has been validated in some PWR static benchmark and the NEACRP PWR transient benchmark cases. However, the NODAL3 code have not yet validated in the transient benchmark cases of a control rod assembly (CR ejection at peripheral core using a full core geometry model, the C1 and C2 cases.  By this research work, the accuracy of the NODAL3 code for one CR ejection or the unsymmetrical group of CRs ejection case can be validated. The calculations by the NODAL3 code have been carried out by the adiabatic method (AM and the improved quasistatic method (IQS. All calculated transient parameters by the NODAL3 code were compared with the reference results by the PANTHER code. The maximum relative difference of 16% occurs in the calculated time of power maximum parameter by using the IQS method, while the relative difference of the AM method is 4% for C2 case.  All calculation results by the NODAL3 code shows there is no systematic difference, it means the neutronic and T/H modules are adopted in the code are considered correct. Therefore, all calculation results by using the NODAL3 code are very good agreement with the reference results. Keywords: nodal method, coupled neutronic and thermal-hydraulic code, PWR, transient case, control rod ejection.   ABSTRAK VALIDASI MODEL GEOMETRI TERAS PENUH PAKET PROGRAM NODAL3 DALAM PROBLEM BENCHMARK GAYUT WAKTU PWR. Paket program kopel neutronik dan termohidraulika (T/H, NODAL3, telah divalidasi dengan beberapa kasus benchmark statis PWR dan kasus benchmark gayut waktu PWR NEACRP.  Akan tetapi, paket program NODAL3 belum divalidasi dalam kasus benchmark gayut waktu akibat penarikan sebuah perangkat batang kendali (CR di tepi teras menggunakan model geometri teras penuh, yaitu kasus C1 dan C2. Dengan penelitian ini, akurasi paket program

  3. Validation of full core geometry model of the NODAL3 code in the PWR transient Benchmark problems

    International Nuclear Information System (INIS)

    T-M Sembiring; S-Pinem; P-H Liem

    2015-01-01

    The coupled neutronic and thermal-hydraulic (T/H) code, NODAL3 code, has been validated in some PWR static benchmark and the NEACRP PWR transient benchmark cases. However, the NODAL3 code have not yet validated in the transient benchmark cases of a control rod assembly (CR) ejection at peripheral core using a full core geometry model, the C1 and C2 cases. By this research work, the accuracy of the NODAL3 code for one CR ejection or the unsymmetrical group of CRs ejection case can be validated. The calculations by the NODAL3 code have been carried out by the adiabatic method (AM) and the improved quasistatic method (IQS). All calculated transient parameters by the NODAL3 code were compared with the reference results by the PANTHER code. The maximum relative difference of 16 % occurs in the calculated time of power maximum parameter by using the IQS method, while the relative difference of the AM method is 4 % for C2 case. All calculation results by the NODAL3 code shows there is no systematic difference, it means the neutronic and T/H modules are adopted in the code are considered correct. Therefore, all calculation results by using the NODAL3 code are very good agreement with the reference results. (author)

  4. Optimization of reload core design for PWR and application to Qinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Shen Wei; Zhongsheng Xie; Banghua Yin

    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. 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

  5. Effect of transplutonium doping on approach to long-life core in uranium-fueled PWR

    Energy Technology Data Exchange (ETDEWEB)

    Peryoga, Yoga; Saito, Masaki; Artisyuk, Vladimir [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors; Shmelev, Anatolii [Moscow Engineering Physics Institute, Moscow (Russian Federation)

    2002-08-01

    The present paper advertises doping of transplutonium isotopes as an essential measure to improve proliferation-resistance properties and burnup characteristics of UOX fuel for PWR. Among them {sup 241}Am might play the decisive role of burnable absorber to reduce the initial reactivity excess while the short-lived nuclides {sup 242}Cm and {sup 244}Cm decay into even plutonium isotopes, thus increasing the extent of denaturation for primary fissile {sup 239}Pu in the course of reactor operation. The doping composition corresponds to one discharged from a current PWR. For definiteness, the case identity is ascribed to atomic percentage of {sup 241}Am, and then the other transplutonium nuclide contents follow their ratio as in the PWR discharged fuel. The case of 1 at% doping to 20% enriched uranium oxide fuel shows the potential of achieving the burnup value of 100 GWd/tHM with about 20% {sup 238}Pu fraction at the end of irradiation. Since so far, americium and curium do not require special proliferation resistance measures, their doping to UOX would assist in introducing nuclear technology in developing countries with simultaneous reduction of accumulated minor actinides stockpiles. (author)

  6. Effect of transplutonium doping on approach to long-life core in uranium-fueled PWR

    International Nuclear Information System (INIS)

    Peryoga, Yoga; Saito, Masaki; Artisyuk, Vladimir

    2002-01-01

    The present paper advertises doping of transplutonium isotopes as an essential measure to improve proliferation-resistance properties and burnup characteristics of UOX fuel for PWR. Among them 241 Am might play the decisive role of burnable absorber to reduce the initial reactivity excess while the short-lived nuclides 242 Cm and 244 Cm decay into even plutonium isotopes, thus increasing the extent of denaturation for primary fissile 239 Pu in the course of reactor operation. The doping composition corresponds to one discharged from a current PWR. For definiteness, the case identity is ascribed to atomic percentage of 241 Am, and then the other transplutonium nuclide contents follow their ratio as in the PWR discharged fuel. The case of 1 at% doping to 20% enriched uranium oxide fuel shows the potential of achieving the burnup value of 100 GWd/tHM with about 20% 238 Pu fraction at the end of irradiation. Since so far, americium and curium do not require special proliferation resistance measures, their doping to UOX would assist in introducing nuclear technology in developing countries with simultaneous reduction of accumulated minor actinides stockpiles. (author)

  7. Computational methods and implementation of the 3-D PWR core dynamics SIMTRAN code for online surveillance and prediction

    International Nuclear Information System (INIS)

    Aragones, J.M.; Ahnert, C.

    1995-01-01

    New computational methods have been developed in our 3-D PWR core dynamics SIMTRAN code for online surveillance and prediction. They improve the accuracy and efficiency of the coupled neutronic-thermalhydraulic solution and extend its scope to provide, mainly, the calculation of: the fission reaction rates at the incore mini-detectors; the responses at the excore detectors (power range); the temperatures at the thermocouple locations; and the in-vessel distribution of the loop cold-leg inlet coolant conditions in the reflector and core channels, and to the hot-leg outlets per loop. The functional capabilities implemented in the extended SIMTRAN code for online utilization include: online surveillance, incore-excore calibration, evaluation of peak power factors and thermal margins, nominal update and cycle follow, prediction of maneuvers and diagnosis of fast transients and oscillations. The new code has been installed at the Vandellos-II PWR unit in Spain, since the startup of its cycle 7 in mid-June, 1994. The computational implementation has been performed on HP-700 workstations under the HP-UX Unix system, including the machine-man interfaces for online acquisition of measured data and interactive graphical utilization, in C and X11. The agreement of the simulated results with the measured data, during the startup tests and first months of actual operation, is well within the accuracy requirements. The performance and usefulness shown during the testing and demo phase, to be extended along this cycle, has proved that SIMTRAN and the man-machine graphic user interface have the qualities for a fast, accurate, user friendly, reliable, detailed and comprehensive online core surveillance and prediction

  8. Treatment of core components from nuclear power plants with PWR and BWR reactors - 16043

    International Nuclear Information System (INIS)

    Viermann, Joerg; Friske, Andreas; Radzuweit, Joerg

    2009-01-01

    During operation of a Nuclear Power Plant components inside the RPV get irradiated. Irradiation has an effect on physical properties of these components. Some components have to be replaced after certain neutron doses or respectively after a certain operating time of the plant. Such components are for instance water channels and control rods from Boiling Water Reactors (BWR) or control elements, poisoning elements and flow restrictors from Pressurized Water Reactors (PWR). Most of these components are stored in the fuel pool for a certain time after replacement. Then they have to be packaged for further treatment or for disposal. More than 25 years ago GNS developed a system for disposal of irradiated core components which was based on a waste container suitable for transport, storage and disposal of Intermediate Level Waste (ILW), the so-called MOSAIK R cask. The MOSAIK R family of casks is subject of a separate presentation at the ICEM 09 conference. Besides the MOSAIK R cask the treatment system developed by GNS comprised underwater shears to cut the components to size as well as different types of equipment to handle the components, the shears and the MOSAIK R casks in the fuel pool. Over a decade of experience it showed that this system although effective needed improvement for BWR plants where many water channels and control rods had to be replaced after a certain operating time. Because of the large numbers of components the time period needed to cut the components in the pool had a too big influence on other operational work like rearranging of fuel assemblies in the pool. The system was therefore further developed and again a suitable cask was the heart of the solution. GNS developed the type MOSAIK R 80 T, a cask that is capable to ship the unsegmented components with a length of approx. 4.5 m from the Power plants to an external treatment centre. This treatment centre consisting of a hot cell installation with a scrap shear, super-compactor and a heavy

  9. Numerical simulation of the insulation material transport to a PWR core under loss of coolant accident conditions

    International Nuclear Information System (INIS)

    Höhne, Thomas; Grahn, Alexander; Kliem, Sören; Rohde, Ulrich; Weiss, Frank-Peter

    2013-01-01

    Highlights: ► Detailed results of a numerical simulation of the insulation material transport to a PWR core are shown. ► The spacer grid is modeled as a strainer which completely retains the insulation material carried by coolant. ► The CFD calculations showed that the fibers at the upper spacer grid plane are not uniformly distributed. ► Furthermore the pressure loss does not exceed a critical limit. ► The PWR core coolablity can be guaranteed all the time during the transient. -- Abstract: In 1992, strainers on the suction side of the ECCS pumps in Barsebäck NPP Unit 2 became partially clogged with mineral wool because after a safety valve opened the steam impinged on thermally insulated equipment and released mineral wool. This event pointed out that strainer clogging is an issue in the course of a loss-of-coolant accident. Modifications of the insulation material, the strainer area and mesh size were carried out in most of the German NPPs. Moreover, back flushing procedures to remove the mineral wool from the strainers and differential pressure measurements were implemented to assure the performance of emergency core cooling during the containment sump recirculation mode. Nevertheless, it cannot be completely ruled out, that a limited amount of small fractions of the insulation material is transported into the RPV. During a postulated cold leg LOCA with hot leg ECC injection, the fibers enter the upper plenum and can accumulate at the fuel element spacer grids, preferably at the uppermost grid level. This effect might affect the ECC flow into the core and could result in degradation of core cooling. It was the aim of the numerical simulations presented to study where and how many mineral wool fibers are deposited at the upper spacer grid. The 3D, time dependent, multi-phase flow problem was modeled applying the CFD code ANSYS CFX. The CFD calculation does not yet include steam production in the core and also does not include re-suspension of the

  10. Comparison of computer codes relative to the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Dunbar, I.; Gauvain, J.; Ricchena, R.

    1986-02-01

    The present study concerns a comparative exercise, performed within the framework of the Commission of the European Communities, of the computer codes (AEROSISM-M, UK; AEROSOLS/BI, France; CORRAL-2, CEC and NAUA Mod5, Germany) used in order to assess the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

  11. PANTHER solution to the NEA-NSC 3-D PWR core transient benchmark. Uncontrolled withdrawal of control rods at zero power

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C.

    1994-10-01

    This report contains the results of PANTHER calculations for the ``NEA-NSC 3-D PWR Core Transient Benchmark: Uncontrolled Withdrawal of Control Rods at Zero Power``. PANTHER was able to model the benchmark problems without modifications to the code. All the calculations were performed in 3-D. (orig.).

  12. PANTHER solution to the NEA-NSC 3-D PWR core transient benchmark. Uncontrolled withdrawal of control rods at zero power

    International Nuclear Information System (INIS)

    Kuijper, J.C.

    1994-10-01

    This report contains the results of PANTHER calculations for the ''NEA-NSC 3-D PWR Core Transient Benchmark: Uncontrolled Withdrawal of Control Rods at Zero Power''. PANTHER was able to model the benchmark problems without modifications to the code. All the calculations were performed in 3-D. (orig.)

  13. Comparison of computer codes relative to the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Bunz, H.; Dunbar, I.; Gauvain, J.; Ricchena, R.

    1986-01-01

    The present study concerns a comparative exercise, performed within the framework of the Commission of the European Communities, of the computer codes (AEROSIM-M, UK; AEROSOLS/B1, France; CORRAL-2, CEC and NAUA Mod5, Germany) used in order to assess the aerosol behavior in the reactor containment building during severe core damage accidents in a PWR. Topics considered in this paper include aerosols, containment buildings, reactor safety, fission product release, reactor cores, meltdown, and monitoring

  14. Application of perturbation theory to sensitivity calculations of PWR type reactor cores using the two-channel model

    International Nuclear Information System (INIS)

    Oliveira, A.C.J.G. de.

    1988-12-01

    Sensitivity calculations are very important in design and safety of nuclear reactor cores. Large codes with a great number of physical considerations have been used to perform sensitivity studies. However, these codes need long computation time involving high costs. The perturbation theory has constituted an efficient and economical method to perform sensitivity analysis. The present work is an application of the perturbation theory (matricial formalism) to a simplified model of DNB (Departure from Nucleate Boiling) analysis to perform sensitivity calculations in PWR cores. Expressions to calculate the sensitivity coefficients of enthalpy and coolant velocity with respect to coolant density and hot channel area were developed from the proposed model. The CASNUR.FOR code to evaluate these sensitivity coefficients was written in Fortran. The comparison between results obtained from the matricial formalism of perturbation theory with those obtained directly from the proposed model makes evident the efficiency and potentiality of this perturbation method for nuclear reactor cores sensitivity calculations (author). 23 refs, 4 figs, 7 tabs

  15. An assessment of the radiological consequences of releases to groundwater following a core-melt accident at the Sizewell PWR

    International Nuclear Information System (INIS)

    Maul, P.R.

    1984-03-01

    In the extremely unlikely event of a degraded core accident at the proposed Sizewell PWR it is theoretically possible for the core to melt through the containment, after which activity could enter groundwater directly or as a result of subsequent leaching of the core in the ground. The radiological consequences of such an event are analysed and compared with the analysis undertaken by the NRPB for the corresponding releases to atmosphere. It is concluded that the risks associated with the groundwater route are much less important than those associated with the atmospheric route. The much longer transport times in the ground compared with those in the atmosphere enable countermeasures to be taken, if necessary, to restrict doses to members of the public to very low levels in the first few years following the accident. The entry of long-lived radionuclides into the sea over very long timescales results in the largest contribution to population doses, but these are delivered at extremely low dose rates which would be negligible compared with background exposure. (author)

  16. Development, verification and validation of an FPGA-based core heat removal protection system for a PWR

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Yichun, E-mail: ycwu@xmu.edu.cn [College of Energy, Xiamen University, Xiamen 361102 (China); Shui, Xuanxuan, E-mail: 807001564@qq.com [College of Energy, Xiamen University, Xiamen 361102 (China); Cai, Yuanfeng, E-mail: 1056303902@qq.com [College of Energy, Xiamen University, Xiamen 361102 (China); Zhou, Junyi, E-mail: 1032133755@qq.com [College of Energy, Xiamen University, Xiamen 361102 (China); Wu, Zhiqiang, E-mail: npic_wu@126.com [State Key Laboratory of Reactor System Design Technology, Nuclear Power Institute of China, Chengdu 610041 (China); Zheng, Jianxiang, E-mail: zwu@xmu.edu.cn [College of Energy, Xiamen University, Xiamen 361102 (China)

    2016-05-15

    Highlights: • An example on life cycle development process and V&V on FPGA-based I&C is presented. • Software standards and guidelines are used in FPGA-based NPP I&C system logic V&V. • Diversified FPGA design and verification languages and tools are utilized. • An NPP operation principle simulator is used to simulate operation scenarios. - Abstract: To reach high confidence and ensure reliability of nuclear FPGA-based safety system, life cycle processes of discipline specification and implementation of design as well as regulations verification and validation (V&V) are needed. A specific example on how to conduct life cycle development process and V&V on FPGA-based core heat removal (CHR) protection system for CPR1000 pressure water reactor (PWR) is presented in this paper. Using the existing standards and guidelines for life cycle development and V&V, a simplified FPGA-based CHR protection system for PWR has been designed, implemented, verified and validated. Diversified verification and simulation languages and tools are used by the independent design team and the V&V team. In the system acceptance testing V&V phase, a CPR1000 NPP operation principle simulator (OPS) model is utilized to simulate normal and abnormal operation scenarios, and provide input data to the under-test FPGA-based CHR protection system and a verified C code CHR function module. The evaluation results are applied to validate the under-test FPGA-based CHR protection system. The OPS model operation outputs also provide reasonable references for the tests. Using an OPS model in the system acceptance testing V&V is cost-effective and high-efficient. A dedicated OPS, as a commercial-off-the-shelf (COTS) item, would contribute as an important tool in the V&V process of NPP I&C systems, including FPGA-based and microprocessor-based systems.

  17. Generalized Thermohydraulics Module GENFLO for Combining With the PWR Core Melting Model, BWR Recriticality Neutronics Model and Fuel Performance Model

    International Nuclear Information System (INIS)

    Miettinen, Jaakko; Hamalainen, Anitta; Pekkarinen, Esko

    2002-01-01

    Thermal hydraulic simulation capability for accident conditions is needed at present in VTT in several programs. Traditional thermal hydraulic models are too heavy for simulation in the analysis tasks, where the main emphasis is the rapid neutron dynamics or the core melting. The GENFLO thermal hydraulic model has been developed at VTT for special applications in the combined codes. The basic field equations in GENFLO are for the phase mass, the mixture momentum and phase energy conservation equations. The phase separation is solved with the drift flux model. The basic variables to be solved are the pressure, void fraction, mixture velocity, gas enthalpy, liquid enthalpy, and concentration of non-condensable gas fractions. The validation of the thermohydraulic solution alone includes large break LOCA reflooding experiments and in specific for the severe accident conditions QUENCH tests. In the recriticality analysis the core neutronics is simulated with a two-dimensional transient neutronics code TWODIM. The recriticality with one rapid prompt peak is expected during a severe accident scenario, where the control rods have been melted and ECCS reflooding is started after the depressurization. The GENFLO module simulates the BWR thermohydraulics in this application. The core melting module has been developed for the real time operator training by using the APROS engineering simulators. The core heatup, oxidation, metal and fuel pellet relocation and corium pool formation into the lower plenum are calculated. In this application the GENFLO model simulates the PWR vessel thermohydraulics. In the fuel performance analysis the fuel rod transient behavior is simulated with the FRAPTRAN code. GENFLO simulates the subchannel around a single fuel rod and delivers the heat transfer on the cladding surface for the FRAPTRAN. The transient boundary conditions for the subchannel are transmitted from the system code for operational transient, loss of coolant accidents and

  18. Calculation of local flow conditions in the lower core of a PWR with code-Saturne

    International Nuclear Information System (INIS)

    Fournier, Y.

    2003-01-01

    In order to better understand the stresses to which fuel rods are subjected, we need to improve our knowledge of the fluid flow inside the core. A code specialized for calculations in tube bundles is used to calculate the flow inside the whole of the core, with a resolution at the assembly level. Still, it is necessary to obtain realistic entry conditions, and these depend on the flow in the downcomer and lower plenum. Also, the flow in the first stages of the core features 4 incoming jets per assembly, and requires a resolution much finer than that used for the whole core calculation. A series of calculations are thus run with our incompressible Navier-Stokes solver, Code-Saturne, using a classical Ranse turbulence model. The first calculations involve a detailed geometry, including part of the cold legs, downcomer, lower plenum, and lower core of a pressurized water reactor. The level of detail includes most obstacles below the core. The lower core plate, being pierced with close to 800 holes, cannot be realistically represented within a practical mesh size, so that a head loss model is used. The lower core itself requiring even more detail is also represented with head losses. We make full use of Code-Saturne's non conforming mesh possibilities to represent a complex geometry, being careful to retain a good mesh quality. Starting just under the lower core, the mesh is aligned with fuel rod assemblies, so that different types of assemblies can be represented through different head loss coefficients. These calculations yield steady-state or near steady-state results, which are compared to experimental data, and should be sufficient to yield realistic entry conditions for full core calculations at assembly width resolution, and beyond those mechanical strain calculations. We are also interested in more detailed flow conditions and fluctuations in the lower core area, so as to better quantify vibrational input. This requires a much higher resolution, which we limit

  19. Study and analysis on the flow induced vibration of the core barrel of PWR

    International Nuclear Information System (INIS)

    Yao Weida; Shi Guolin; Jiang Nanyan; Peng YongYong; Zhang Huijun; Wang Yufen; Xie Yongcheng; Guo Chunhua; Shen Qinping

    1989-01-01

    The deduction of the resemblance criterion and the design of the test model by applying flow-solid coupling theory are described. The model analysis of a core barrel both in the air and stationary water were performed in a 1:10 model, thus obtaining the dynamic characteristic. In a 1:5 reactor model with a hydraulic closed loop, the inner structure and support were modeled for performing hydraulic closed loop, the inner structure and support were modeled for performing hydraulic vibration test of the core barrel. The flow induced pulse pressure of the core barrel and corresponding response were obtained by using miniature pressure capsule, strain gauge and accelerometer. Power spectrum, correlation functions, transfer function and amplitudes under different flow velocities were calculated. The hydraulic vibration test shows that the core barrel will be in safety during its 30-year life time

  20. Improvement of Axial Reflector Cross Section Generation Model for PWR Core Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Cheon Bo; Lee, Kyung Hoon; Cho, Jin Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    This paper covers the study for improvement of axial reflector XS generation model. In the next section, the improved 1D core model is represented in detail. Reflector XS generated by the improved model is compared to that of the conventional model in the third section. Nuclear design parameters generated by these two XS sets are also covered in that section. Significant of this study is discussed in the last section. Two-step procedure has been regarded as the most practical approach for reactor core designs because it offers core design parameters quite rapidly within acceptable range. Thus this approach is adopted for SMART (System-integrated Modular Advanced Reac- Tor) core design in KAERI with the DeCART2D1.1/ MASTER4.0 (hereafter noted as DeCART2D/ MASTER) code system. Within the framework of the two-step procedure based SMART core design, various researches have been studied to improve the core design reliability and efficiency. One of them is improvement of reflector cross section (XS) generation models. While the conventional FA/reflector two-node model used for most core designs to generate reflector XS cannot consider the actual configuration of fuel rods that intersect at right angles to axial reflectors, the revised model reflects the axial fuel configuration by introducing the radially simplified core model. The significance of the model revision is evaluated by observing HGC generated by DeCART2D, reflector XS, and core design parameters generated by adopting the two models. And it is verified that about 30 ppm CBC error can be reduced and maximum Fq error decreases from about 6 % to 2.5 % by applying the revised model. Error of AO and axial power shapes are also reduced significantly. Therefore it can be concluded that the simplified 1D core model improves the accuracy of the axial reflector XS and leads to the two-step procedure reliability enhancement. Since it is hard for core designs to be free from the two-step approach, it is necessary to find

  1. Influence of fuel vibration on PWR neutron noise associated with core barrel motion

    International Nuclear Information System (INIS)

    Sweeney, F.J.; March-Leuba, J.

    1984-01-01

    Ex-core neutron detector noise has been utilized to monitor core support barrel (CSB) vibrations. In order to observe long-term changes, noise signals at Sequoyah-1 were monitored continuously during the whole first fuel cycle and part of the second cycle. Results suggest that neutron noise measurements performed infrequently may not provide adequate surveillance of the CSB because it may be difficult to separate noise amplitude changes due solely to CSB motion from changes caused by fuel motion and burnup

  2. 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)

  3. MC21 Monte Carlo analysis of the Hoogenboom-Martin full-core PWR benchmark problem - 301

    International Nuclear Information System (INIS)

    Kelly, D.J.; Sutton, Th.M.; Trumbull, T.H.; Dobreff, P.S.

    2010-01-01

    At the 2009 American Nuclear Society Mathematics and Computation conference, Hoogenboom and Martin proposed a full-core PWR model to monitor the improvement of Monte Carlo codes to compute detailed power density distributions. This paper describes the application of the MC21 Monte Carlo code to the analysis of this benchmark model. With the MC21 code, we obtained detailed power distributions over the entire core. The model consisted of 214 assemblies, each made up of a 17x17 array of pins. Each pin was subdivided into 100 axial nodes, thus resulting in over seven million tally regions. Various cases were run to assess the statistical convergence of the model. This included runs of 10 billion and 40 billion neutron histories, as well as ten independent runs of 4 billion neutron histories each. The 40 billion neutron-history calculation resulted in 43% of all regions having a 95% confidence level of 2% or less implying a relative standard deviation of 1%. Furthermore, 99.7% of regions having a relative power density of 1.0 or greater have a similar confidence level. We present timing results that assess the MC21 performance relative to the number of tallies requested. Source convergence was monitored by analyzing plots of the Shannon entropy and eigenvalue versus active cycle. We also obtained an estimate of the dominance ratio. Additionally, we performed an analysis of the error in an attempt to ascertain the validity of the confidence intervals predicted by MC21. Finally, we look forward to the prospect of full core 3-D Monte Carlo depletion by scoping out the required problem size. This study provides an initial data point for the Hoogenboom-Martin benchmark model using a state-of-the-art Monte Carlo code. (authors)

  4. Modelling of core protection and monitoring system for PWR nuclear power plant simulator

    International Nuclear Information System (INIS)

    Jung Kun Lee; Byoung Sung Han

    1997-01-01

    A nuclear power plant simulator was developed for Younggwang units 3 and 4 nuclear power plant (YGN Nos 3 and 4) in Korea; it has been in operation on training center since November 1996. The core protection calculator (CPC) and the core operating limit supervisory system (COLSS) for the simulator were also developed. The CPC is a digital computer-based core protection system, which performs on-line calculation of departure from nucleate boiling ratio (DNBR) and local power density (LPD). It initiates reactor trip when the core conditions exceed designated DNBR or LPD limitations. The COLSS is designed to assist operators by implementing the limiting conditions for operations in the technical specifications. With these systems, it is possible to increase capacity factor and safety of nuclear power plants, because the COLSS data can show accurate operation margin to plant operators and the CPC can protect reactor core. In this study, the function of CPC/COLSS is analyzed in detail, and then simulation model for CPC/COLSS is presented based on the function. Compared with the YGN Nos 3 and 4 plant operation data and CEDIPS/COLSS FORTRAN code test results, the predictions with the model show reasonable results. (Author)

  5. Quantification of cost of margin associated with in-core nuclear fuel management for a PWR

    International Nuclear Information System (INIS)

    Kropaczek, D.J.; Turinsky, P.J.

    1989-01-01

    The problem of in-core nuclear fuel management optimization is discussed. The problem is to determine the location of core material, such as the fuel and burnable poisons, so as to minimize (maximize) a stated objective within engineering constraints. Typical objectives include maximization of cycle energy production or discharged fuel exposure, and minimization of power peaking factor or reactor vessel fluence. Constraints include discharge burnup limits and one or more of the possible objectives if not selected as the objective. The optimization problem can be characterized as a large combinatorial problem with nonlinear objective function and constraints, which are likely to be active. The authors have elected to employ the integer Monte Carlo programming method to address this optimization problem because of the just-noted problem characteristics. To evaluate the core physics characteristics as a function of fuel loading pattern, second-order accurate perturbation theory is employed with successive application to improve estimates of the optimum loading pattern. No constraints on fuel movement other than requiring quarter-core symmetry were imposed. In this paper the authors employed this methodology to address a related problem. The problem being addressed can be stated as What is the cost associated with margin? Specifically, they wish to assign some financial value in terms of increased levelized fuel cycle cost associated with an increase in core margin of some type, such as power peaking factor

  6. Design study on PWR-type reduced-moderation light water core. Investigation of core adopting seed-blanket fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Shimada, Shoichiro; Kugo, Teruhiko; Okubo, Tsutomu; Iwamura, Takamichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    As a part of the design study on PWR-type Reduced-Moderation Water Reactors (RMWRs), a light water cooled core with the seed-blanket type fuel assemblies has been investigated. An assembly with seed of 13 layers and blanket of 5 layers was selected by optimization calculations. The core was composed with the 163 assemblies. The following results were obtained by burn-up calculations with the MVP-BURN code; The cycle length is 15 months by 3-batch refueling. The discharge burn-up including the inner blanket is about 25 GWd/t. The conversion ratio is about 1.0. The void reactivity coefficient is about-26.1 pcm/%void at BOC and -21.7pcm%void at EOC. About 10% of MA makes conversion ratio decrease about 0.05 to obtain the same burn-up. The void reactivity coefficient increased significantly and it is necessary to reduce it. FP amount corresponding to about 2 % of total plutonium weight makes reactivity decrease about 0.5 %{delta}k/k and void reactivity coefficient increase, however these changes are within the design margins. Capability of multi-recycling of plutonium was confirmed, using discharged plutonium for 4 cycles, if fissile plutonium of 15.5wt% is used. The conversion ratio increases by about 0.026 with recycling. However, void reactivity coefficient increases and some effort to obtain negative void reactivity coefficient is necessary. (author)

  7. Study for on-line system to identify inadvertent control rod drops in PWR reactors using ex-core detector and thermocouple measures

    International Nuclear Information System (INIS)

    Souza, Thiago J.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C.

    2015-01-01

    Accidental control rod drops event in PWR reactors leads to an unsafe operating condition. It is important to quickly identify the rod to minimize undesirable effects in such a scenario. In this event, there is a distortion in the power distribution and temperature in the reactor core. The goal of this study is to develop an on-line model to identify the inadvertent control rod dropped in PWR reactor. The proposed model is based on physical correlations and pattern recognition of ex-core detector responses and thermocouples measures. The results of the study demonstrated the feasibility of an on-line system, contributing to safer operation conditions and preventing undesirable effects, as its shutdown. (author)

  8. Study for on-line system to identify inadvertent control rod drops in PWR reactors using ex-core detector and thermocouple measures

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Thiago J.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C., E-mail: tsouza@nuclear.ufrj.br, E-mail: canedo@lmp.ufrj.br, E-mail: alessandro@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2015-07-01

    Accidental control rod drops event in PWR reactors leads to an unsafe operating condition. It is important to quickly identify the rod to minimize undesirable effects in such a scenario. In this event, there is a distortion in the power distribution and temperature in the reactor core. The goal of this study is to develop an on-line model to identify the inadvertent control rod dropped in PWR reactor. The proposed model is based on physical correlations and pattern recognition of ex-core detector responses and thermocouples measures. The results of the study demonstrated the feasibility of an on-line system, contributing to safer operation conditions and preventing undesirable effects, as its shutdown. (author)

  9. RELAP5/MOD3.3 Analyses of Core Heatup Prevention Strategy During Extended Station Blackout in PWR

    International Nuclear Information System (INIS)

    Prosek, A.

    2016-01-01

    The accident at the Fukushima Dai-ichi nuclear power plant demonstrated the vulnerability of the plants on the loss of electrical power for several days, so called extended station blackout (SBO). A set of measures have been proposed and implemented in response of the accident at the Fukushima Dai-ichi nuclear power plant. The purpose of the study was to investigate the application of the deterministic safety analysis for core heatup prevention strategy of the extended SBO in pressurized water reactor, lasting 72 h. The prevention strategy selected was water injection into steam generators using turbine driven auxiliary feedwater pump (TD-AFW) or portable water injection pump. Method for assessment of the necessary pump injection flowrate is developed and presented. The necessary injection flowrate to the steam generators is determined from the calculated cumulative water mass injected by the turbine driven auxiliary feedwater pump in the analysed scenarios, when desired normal level is maintained automatically. The developed method allows assessment of the necessary injection flowrates of pump, TD-AFW or portable, for different plant configurations and number of flowrate changes. The RELAP5/MOD3.3 Patch04 computer code and input model of a two-loop pressurized water reactor is used for analyses, assuming different injection start times, flowrates and reactor coolant system losses. Three different reactor coolant system (RCS) coolant loss pathways, with corresponding leakage rate, can be expected in the pressurized water reactor (PWR) during the extended SBO: normal system leakage, reactor coolant pump seal leakage, and RCS coolant loss through letdown relief valve unless automatically isolated or until isolation is procedurally directed. Depressurization of RCS was also considered. In total, six types of RCS coolant loss scenarios were considered. Two cases were defined regarding the operation of the emergency diesel generators. Different delays of the pump

  10. Uncertainty evaluatins of CASMO-3/MASTER system for PWR core neutronics calculations

    International Nuclear Information System (INIS)

    Song, Jae Seung; Kim, Kang Seog; Lee, Kibog; Park, Jin Ha; Zee, Sung Quun

    1996-01-01

    Uncertainties in core neutronic calculations of CASMO-3/MASTER, which is a KAERI developed core nuclear design code system, were evaluated via comparisons with measured data. Comparisons were performed with plant measurement data from one Westinghouse type and one ABB-CE type plant and two Korean standard type plants. The CASMO-3/MASTER capability and levels of accuracy are concluded to be sufficient for the neutronics design including safety related parameters related with reactivity, power distributions, temperature and power coefficients, inverse boron worth and control bank worth

  11. Comparison of european computer codes relative to the aerosol behavior in PWR containment buildings during severe core damage accidents. (Modelling of steam condensation on the particles)

    International Nuclear Information System (INIS)

    Bunz, H.; Dunbar, L.H.; Fermandjian, J.; Lhiaubet, G.

    1987-11-01

    An aerosol code comparison exercise was performed within the framework of the Commission of European Communities (Division of Safety of Nuclear Installations). This exercise, focused on the process of steam condensation onto the aerosols occurring in PWR containment buildings during severe core damage accidents, has allowed to understand the discrepancies between the results obtained. These discrepancies are due, in particular, to whether the curvature effect is modelled or not in the codes

  12. Analysis of subchannel effects and their treatment in average channel PWR core models

    International Nuclear Information System (INIS)

    Cuervo, D.; Ahnert, C.; Aragones, J.M.

    2004-01-01

    Neutronic thermal-hydraulic coupling is meanly made at this moment using whole plant thermal-hydraulic codes with one channel per assembly or quarter of assembly in more detailed cases. To extract safety limits variables a new calculation has to be performed using thermal-hydraulic subchannel codes in an embedded or off-line manner what implies an increase of calculation time. Another problem of this separated analysis of whole core and not channel is that the whole core calculation is not resolving the real problem due to the modification of the variables values by the homogenization process that is carried out to perform the whole core analysis. This process is making that some magnitudes are over or under-predicted causing that the problem that is being solved is not the original one. The purpose of the work that is being developed is to investigate the effects of the averaging process in the results obtained by the whole core analysis and to develop some corrections that may be included in this analysis to obtain results closer to the ones obtained by a detailed subchannel analysis. This paper shows the results obtained for a sample case and the conclusions for future work. (author)

  13. A non-algorithmic approach to the In-core-fuel management problem of a PWR core

    International Nuclear Information System (INIS)

    Kimhy, Y.

    1992-03-01

    The primary objective of a commercial nuclear power plant operation is to produce electricity a low cost while satisfying safety constraints imposed on the operating conditions. Design of a fuel reload cycle for the current generation nuclear power plant represents a multistage process with a series of design decisions taken at various time points. Of these stages, reload core design is an important stage, due to its impact on safety and economic plant performance parameters. Overall. performance of the plant during the power production cycle depends on chosen fresh fuel parameters, as well as specific fuel configuration of the reactor core. The motivation to computerize generation and optimization of fuel reload configurations follows from some reasons: first, reload is performed periodically and requires manipulation of a large amount of data. second, in recent years, more complicated fuel loading patterns were developed and implemented following changes in fuel design and/or operational requirements, such as, longer cycles, advanced burnable poison designs, low leakage loading patterns and reduction of irradiation-induced damage of the pressure vessel. An algorithmic approach to the problem was generally adopted. The nature of the reload design process is a 'heuristic' search performed manually by a fuel manager. The knowledge used by the fuel manager is mostly accumulated experience in reactor physics and core calculations. These features of the problem and the inherent disadvantage of the algorithmic method are the main reasons to explore a non-algorithmic approach for solving the reload configuration problem. Several features of the 'solutions space' ( a collection of acceptable final configurations ) are emphasized in this work: 1) the space contain numerous number of entities (> 25) that are distributed un homogeneously, 2) the lack of a monotonic objective function decrease the probability to find an isolated optimum configuration by depth first search or

  14. ACTRAN: a code for depletion calculations in PWR cores aiming the production of minor actinide for using in ADS

    International Nuclear Information System (INIS)

    Santos, Rubens Souza dos

    2009-01-01

    Despite of the renewed willing to accept nuclear power as a mean of mitigate the climate changing, to deal with the long lived waste still cause some concerning in relation to maintain in safety condition, during so many years. A technological solution to overcome this leg of time is to use a facility that burn these waste, besides to generate electricity. This is the idea built in the accelerator driven systems (ADS). This technology is being though to use some minor actinides (MAs) as fuel. This work presents a program to assess actinide concentrations, aiming a fertile-free fuel to be used in the future ADS technology. For that, use was made of a numerical code to solve the steady-state multigroup diffusion equation 3D to calculate the neutron fluxes, coupled it with a new code to solve, also numerically, depletion equations, named ACTRAN code. This paper shows the simulation of a PWR core during the residence time of the nuclear fuel, for three years, and after, for almost four hundred years, to assess the MAs production. The results show some insight in the best management to get a minimum amount of some MAs to use in the future generations of ADS. (author)

  15. Advanced methods for the study of PWR cores; Les methodes d'etudes avancees pour les coeurs de REP

    Energy Technology Data Exchange (ETDEWEB)

    Lambert, M.; Salvatores, St.; Ferrier, A. [Electricite de France (EDF), Service Etudes et Projets Thermiques et Nucleaires, 92 - Courbevoie (France); Pelet, J.; Nicaise, N.; Pouliquen, J.Y.; Foret, F. [FRAMATOME ANP, 92 - Paris La Defence (France); Chauliac, C. [CEA Saclay, Dir. de l' Energie Nucleaire (DEN), 91 - Gif sur Yvette (France); Johner, J. [CEA Cadarache, Dept. de Recherches sur la Fusion Controlee (DRFC), 13 - Saint Paul lez Durance (France); Cohen, Ch

    2003-07-01

    This document gathers the transparencies presented at the 6. technical session of the French nuclear energy society (SFEN) in October 2003. The transparencies of the annual meeting are presented in the introductive part: 1 - status of the French nuclear park: nuclear energy results, management of an exceptional climatic situation: the heat wave of summer 2003 and the power generation (J.C. Barral); 2 - status of the research on controlled thermonuclear fusion (J. Johner). Then follows the technical session about the advanced methods for the study of PWR reactor cores: 1 - the evolution approach of study methodologies (M. Lambert, J. Pelet); 2 - the point of view of the nuclear safety authority (D. Brenot); 3 - the improved decoupled methodology for the steam pipe rupture (S. Salvatores, J.Y. Pouliquen); 4 - the MIR method for the pellet-clad interaction (renovated IPG methodology) (E. Baud, C. Royere); 5 - the improved fuel management (IFM) studies for Koeberg (C. Cohen); 6 - principle of the methods of accident study implemented for the European pressurized reactor (EPR) (F. Foret, A. Ferrier); 7 - accident studies with the EPR, steam pipe rupture (N. Nicaise, S. Salvatores); 8 - the co-development platform, a new generation of software tools for the new methodologies (C. Chauliac). (J.S.)

  16. Computational simulation of natural convection of a molten core in lower head of a PWR pressure vessel

    International Nuclear Information System (INIS)

    Vieira, Camila Braga; Romero, Gabriel Alves; Jian Su

    2010-01-01

    Computational simulation of natural convection in a molten core during a hypothetical severe accident in the lower head of a typical PWR pressure vessel was performed for two-dimensional semi-circular geometry with isothermal walls. Transient turbulent natural convection heat transfer of a fluid with uniformly distributed volumetric heat generation rate was simulated by using a commercial computational fluid dynamics software ANSYS CFX 12.0. The Boussinesq model was used for the buoyancy effect generated by the internal heat source in the flow field. The two-equation k-ω based SST (Shear Stress Transport) turbulence model was used to mould the turbulent stresses in the Reynolds-Average Navier-Stokes equations (RANS). Two Prandtl numbers, 6:13 and 7:0, were considered. Five Rayleigh numbers were simulated for each Prandtl number used (109, 1010, 1011, 1012, and 1013). The average Nusselt numbers on the bottom surface of the semicircular cavity were in excellent agreement with Mayinger et al. (1976) correlation and only at Ra = 109 the average Nusselt number on the top flat surface was in agreement with Mayinger et al. (1976) and Kulacki and Emara (1975) correlations. (author)

  17. Simulation of single-phase rod bundle flow. Comparison between CFD-code ESTET, PWR core code THYC and experimental results

    International Nuclear Information System (INIS)

    Mur, J.; Larrauri, D.

    1998-07-01

    Computer simulation of flow in configurations close to pressurized water reactor (PWR) geometry is of great interest for Electricite de France (EDF). Although simulation of the flow through a whole PWR core with an all purpose CFD-code is not yet achievable, such a tool cna be quite useful to perform numerical experiments in order to try and improve the modeling introduced in computer codes devoted to reactor core thermal-hydraulic analysis. Further to simulation in small bare rod bundle configurations, the present study is focused on the simulation, with CFD-code ESTET and PWR core code THYC, of the flow in the experimental configuration VATICAN-1. ESTET simulation results are compared on the one hand to local velocity and concentration measurements, on the other hand with subchannel averaged values calculated by THYC. As far as the comparison with measurements is concerned, ESTET results are quite satisfactory relatively to available experimental data and their uncertainties. The effect of spacer grids and the prediction of the evolution of an unbalanced velocity profile seem to be correctly treated. As far as the comparison with THYC subchannel averaged values is concerned, the difficulty of a direct comparison between subchannel averaged and local values is pointed out. ESTET calculated local values are close to experimental local values. ESTET subchannel averaged values are also close to THYC calculation results. Thus, THYC results are satisfactory whereas their direct comparison to local measurements could show some disagreement. (author)

  18. Emotional learning based intelligent controller for a PWR nuclear reactor core during load following operation

    International Nuclear Information System (INIS)

    Khorramabadi, Sima Seidi; Boroushaki, Mehrdad; Lucas, Caro

    2008-01-01

    The design and evaluation of a novel approach to reactor core power control based on emotional learning is described. The controller includes a neuro-fuzzy system with power error and its derivative as inputs. A fuzzy critic evaluates the present situation, and provides the emotional signal (stress). The controller modifies its characteristics so that the critic's stress is reduced. Simulation results show that the controller has good convergence and performance robustness characteristics over a wide range of operational parameters

  19. Simulation study on insoluble granular corrosion products deposited in PWR core

    International Nuclear Information System (INIS)

    Yang Xu; Zhou Tao; Ru Xiaolong; Lin Daping; Fang Xiaolu

    2014-01-01

    In the operation of reactor, such as fuel rods, reactor vessel internals etc. will be affected by corrosion erosion of high pressure coolant. It will produce many insoluble corrosion products. The FLUENT software is adopted to simulate insoluble granular corrosion products deposit distribution in the reactor core. The fluid phase uses the standard model to predict the flow field in the channel and forecast turbulence variation in the near-wall region. The insoluble granular corrosion products use DPM (Discrete Phase Model) to track the trajectory of the particles. The discrete phase model in FLUENT follows the Euler-Lagrange approach. The fluid phase is treated as a continuum by solving the Navier-Stokes equations, while the dispersed phase is solved by tracking a large number of particles through the calculated flow field. Through the study found, Corrosion products particles form high concentration area near the symmetry, and the entrance section of the corrosion products particles concentration is higher than export section. Corrosion products particles deposition attached on large area for the entrance of the cladding, this will change the core neutron flux distribution and the thermal conductivity of cladding material, and cause core axial offset anomaly (AOA). Corrosion products particles dot deposit in the outlet of cladding, which can lead to pitting phenomenon in a sheath. Pitting area will cause deterioration of heat transfer, destroy the cladding integrity. In view of the law of corrosion products deposition and corrosion characteristics of components in the reactor core. this paper proposes regular targeted local cleanup and other mitigation measures. (authors)

  20. Design and fuel management of PWR cores to optimize the once-through fuel cycle

    International Nuclear Information System (INIS)

    Fujita, E.K.; Driscoll, M.J.; Lanning, D.D.

    1978-08-01

    The once-through fuel cycle has been analyzed to see if there are substantial prospects for improved uranium ore utilization in current light water reactors, with a specific focus on pressurized water reactors. The types of changes which have been examined are: (1) re-optimization of fuel pin diameter and lattice pitch, (2) axial power shaping by enrichment gradation in fresh fuel, (3) use of 6-batch cores with semi-annual refueling, (4) use of 6-batch cores with annual refueling, hence greater extended (approximately doubled) burnup, (5) use of radial reflector assemblies, (6) use of internally heterogeneous cores (simple seed/blanket configurations), (7) use of power/temperature coastdown at the end of life to extend burnup, (8) use of metal or diluted oxide fuel, (9) use of thorium, and (10) use of isotopically separated low sigma/sub a/ cladding material. State-of-the-art LWR computational methods, LEOPARD/PDQ-7/FLARE-G, were used to investigate these modifications

  1. Streamlined analysis technique for the evaluation of pellet clad interaction in PWR reload cores

    International Nuclear Information System (INIS)

    Beard, Ch.; Morita, T.; Brown, J.

    2007-01-01

    For some applications, an analysis is required to explicitly demonstrate that fuel failure due to pellet-clad interaction (PCI) is prevented by the core limits and the protection system for both Condition I (normal operation) operation and for Condition II (events of moderate frequency) events. This analysis needs to address the entire range of normal operation allowed by the Technical Specifications and all Condition II transients. The obvious approach which has been utilized for many years is a simulation of normal operation power maneuvers followed by explicit Condition II transients as a function of key core parameters. This is a sampling approach and has concerns about the overall coverage of the potential space. An alternative approach is the 3D FAC power distribution analysis methodology that was based upon the Westinghouse Relaxed Axial Offset Control Strategy (RAOC) evaluation process. The 3D FAC methodology uses a parametric representation of variables affecting the power distributions, defining a grid mesh over a space of Condition I and Condition II parameters. The operation space is defined by a power range, temperature range, rod position range, axial offset range, core protection limits and representative xenon distributions. Then the 3D FAC evaluation consists of systematically calculating the 3D power distribution and margin to the core and fuel limits for each mesh point of this multi-dimensional space. The PCI margin is obtained by the comparison of the 3D power distributions over the Condition II space and the 3D maximum allowed power, which is dependent on the fuel rod history. The fuel history model utilizes the power history developed in the 3-dimensional nuclear analysis code to define local powers for the specified fuel rods to be analyzed. It tracks the rod history and provides the maximum allowed power for the point. This model is appropriate for base load operation, extended reduced power operation, return to power operation and

  2. Streamlined analysis technique for the evaluation of pellet clad interaction in PWR reload cores

    Energy Technology Data Exchange (ETDEWEB)

    Beard, Ch.; Morita, T.; Brown, J. [Westinghouse Electric Company, LLC, Nuclear Fuel Div., Pittsburgh, PA (United States)

    2007-07-01

    For some applications, an analysis is required to explicitly demonstrate that fuel failure due to pellet-clad interaction (PCI) is prevented by the core limits and the protection system for both Condition I (normal operation) operation and for Condition II (events of moderate frequency) events. This analysis needs to address the entire range of normal operation allowed by the Technical Specifications and all Condition II transients. The obvious approach which has been utilized for many years is a simulation of normal operation power maneuvers followed by explicit Condition II transients as a function of key core parameters. This is a sampling approach and has concerns about the overall coverage of the potential space. An alternative approach is the 3D FAC power distribution analysis methodology that was based upon the Westinghouse Relaxed Axial Offset Control Strategy (RAOC) evaluation process. The 3D FAC methodology uses a parametric representation of variables affecting the power distributions, defining a grid mesh over a space of Condition I and Condition II parameters. The operation space is defined by a power range, temperature range, rod position range, axial offset range, core protection limits and representative xenon distributions. Then the 3D FAC evaluation consists of systematically calculating the 3D power distribution and margin to the core and fuel limits for each mesh point of this multi-dimensional space. The PCI margin is obtained by the comparison of the 3D power distributions over the Condition II space and the 3D maximum allowed power, which is dependent on the fuel rod history. The fuel history model utilizes the power history developed in the 3-dimensional nuclear analysis code to define local powers for the specified fuel rods to be analyzed. It tracks the rod history and provides the maximum allowed power for the point. This model is appropriate for base load operation, extended reduced power operation, return to power operation and

  3. A seismic analysis of Korean standard PWR fuels under transition core conditions

    International Nuclear Information System (INIS)

    Kim, Hyeong Koo; Park, Nam Kyu; Jang, Young Ki; Kim, Jae Ik; Kim, Kyu Tae

    2005-01-01

    The PLUS7 fuel is developed to achieve higher thermal performance, burnup and more safety margin than the conventional fuel used in the Korean Standard Nuclear Plants (KSNPs) and to sustain structural integrity under increased seismic requirement in Korea. In this study, a series of seismic analysis have been performed in order to evaluate the structural integrity of fuel assemblies associated with seismic loads in the KSNPs under transition core conditions replacing the Guardian fuel, which is a resident fuel in the KSNP reactors, with the PLUS7 fuel. For the analysis, transition core seismic models have been developed, based on the possible fuel loading patterns. And the maximum impact forces on the spacer grid and various stresses acting on the fuel components have been evaluated and compared with the through-grid strength of spacer grids and the stress criteria specified in the ASME code for each fuel component, respectively. Then three noticeable parameters regarding as important parameters governing fuel assembly dynamic behavior are evaluated to clarify their effects on the fuel impact and stress response. As a result of the study, it has been confirmed that both the PLUS7 and the Guardian fuel sustain their structural integrity under the transition core condition. And when the damping ratio is constant, increasing the natural frequency of fuel assembly results in a decrease in impact force. The fuel assembly flexural stiffness has an effect increasing the stress of fuel assembly, but not the impact force. And the spacer grid stiffness is directly related with the impact force response. (author)

  4. Interaction between core analysis methodology and nuclear design: some PWR examples

    International Nuclear Information System (INIS)

    Rothleder, B.M.; Eich, W.J.

    1982-01-01

    The interaction between core analysis methodology and nuclear design is exemplified by PSEUDAX, a major improvement related to the Advanced Recycle methodology program (ARMP) computer code system, still undergoing development by the Electric Power Research Institute. The mechanism of this interaction is explored by relating several specific nulcear design changes to the demands placed by these changes on the ARMP system, and by examining the meeting of these demands, first within the standard ARMP methodology and then through augmentation of the standard methodology by development of PSEUDAX

  5. PWR Core II blanket fuel disposition recommendation of storage option study

    International Nuclear Information System (INIS)

    Dana, C.M.

    1995-01-01

    After review of the options available for current storage of T Plant Fuel the recommended option is wet storage without the use of chillers. A test has been completed that verifies the maximum temperature reached is below the industrial standard for storage of spent fuel. This option will be the least costly and still maintain the fuel in a safe environment. The options that were evaluated included dry storage with and without chillers, and wet storage with and without chillers. Due to the low decay heat of the Shippingport Core II Blanket fuel assemblies the fuel pool temperature will not exceed 100 deg. F

  6. Thermal-hydraulic analysis of PWR core including intermediate flow mixers with the THYC code

    International Nuclear Information System (INIS)

    Mur, J.; Meignin, J.C.

    1997-07-01

    Departure from nucleate boiling (DNB) is one of the major limiting factors of pressurized water reactors (PWRs). Safety requires that occurrence of DNB should be precluded under normal or incidental operating conditions. The thermal-hydraulic THYC code developed by EDF is described. The code is devoted to heat and mass transfer in nuclear components. Critical Heat Flux (CHF) is predicted from local thermal-hydraulic parameters such as pressure, mass flow rate, and quality. A three stage methodology to evaluate thermal margins in order to perform standard core design is described. (K.A.)

  7. Thermal-hydraulic analysis of PWR core including intermediate flow mixers with the THYC code

    Energy Technology Data Exchange (ETDEWEB)

    Mur, J. [Electricite de France (EDF), 78 - Chatou (France); Meignin, J.C. [Electricite de France (EDF), 69 - Villeurbanne (France)

    1997-07-01

    Departure from nucleate boiling (DNB) is one of the major limiting factors of pressurized water reactors (PWRs). Safety requires that occurrence of DNB should be precluded under normal or incidental operating conditions. The thermal-hydraulic THYC code developed by EDF is described. The code is devoted to heat and mass transfer in nuclear components. Critical Heat Flux (CHF) is predicted from local thermal-hydraulic parameters such as pressure, mass flow rate, and quality. A three stage methodology to evaluate thermal margins in order to perform standard core design is described. (K.A.) 8 refs.

  8. Investigation of novel inverted NiO@NixCo1-xO core-shell nanoparticles

    Science.gov (United States)

    Hasan, Samiul; Mayanovic, R. A.; Benamara, Mourad

    2018-05-01

    Inverse core-shell nanoparticles, comprised of an antiferromagnetic (AFM) core covered by a ferromagnetic (FM) or ferrimagnetic (FiM) shell, are of current interest due to their different potential application and due to the tunability of their magnetic properties. The antiferromagnetic nature of NiO and high Néel temperature (523 K) makes this material well suited for inverse core-shell nanoparticle applications. Our primary objective in this project has been to synthesize and characterize inverted core-shell nanoparticles (CSNs) comprised of a NiO (AFM) core and a shell consisting of a NixCo1-xO (FiM) compound. The synthesis of the CSNs was made using a two-step process. The NiO nanoparticles were synthesized using a chemical reaction method. Subsequently, the NiO nanoparticles were used to grow the NiO@NixCo1-xO CSNs using our hydrothermal nano-phase epitaxy method. XRD structural characterization shows that the NiO@NixCo1-xO CSNs have the rock salt cubic crystal structure. SEM-EDS data indicates the presence of Co in the CSNs. Magnetic measurements show that the CSNs exhibit AFM/FiM characteristics with a small coercivity field of 30 Oe at 5 K. The field cooled vs zero field cooled hysteresis loop measurements show a magnetization axis shift which is attributed to the exchange bias effect between the AFM NiO core and an FiM NixCo1-xO shell of the CSNs. Our ab initio based calculations of the NixCo1-xO rock salt structure confirm a weak FiM character and a charge transfer insulator property of the compound.

  9. Investigation of novel inverted NiO@NixCo1-xO core-shell nanoparticles

    Directory of Open Access Journals (Sweden)

    Samiul Hasan

    2018-05-01

    Full Text Available Inverse core-shell nanoparticles, comprised of an antiferromagnetic (AFM core covered by a ferromagnetic (FM or ferrimagnetic (FiM shell, are of current interest due to their different potential application and due to the tunability of their magnetic properties. The antiferromagnetic nature of NiO and high Néel temperature (523 K makes this material well suited for inverse core-shell nanoparticle applications. Our primary objective in this project has been to synthesize and characterize inverted core-shell nanoparticles (CSNs comprised of a NiO (AFM core and a shell consisting of a NixCo1-xO (FiM compound. The synthesis of the CSNs was made using a two-step process. The NiO nanoparticles were synthesized using a chemical reaction method. Subsequently, the NiO nanoparticles were used to grow the NiO@NixCo1-xO CSNs using our hydrothermal nano-phase epitaxy method. XRD structural characterization shows that the NiO@NixCo1-xO CSNs have the rock salt cubic crystal structure. SEM-EDS data indicates the presence of Co in the CSNs. Magnetic measurements show that the CSNs exhibit AFM/FiM characteristics with a small coercivity field of 30 Oe at 5 K. The field cooled vs zero field cooled hysteresis loop measurements show a magnetization axis shift which is attributed to the exchange bias effect between the AFM NiO core and an FiM NixCo1-xO shell of the CSNs. Our ab initio based calculations of the NixCo1-xO rock salt structure confirm a weak FiM character and a charge transfer insulator property of the compound.

  10. INSIGHT: an integrated scoping analysis tool for in-core fuel management of PWR

    International Nuclear Information System (INIS)

    Yamamoto, Akio; Noda, Hidefumi; Ito, Nobuaki; Maruyama, Taiji.

    1997-01-01

    An integrated software tool for scoping analysis of in-core fuel management, INSIGHT, has been developed to automate the scoping analysis and to improve the fuel cycle cost using advanced optimization techniques. INSIGHT is an interactive software tool executed on UNIX based workstations that is equipped with an X-window system. INSIGHT incorporates the GALLOP loading pattern (LP) optimization module that utilizes hybrid genetic algorithms, the PATMAKER interactive LP design module, the MCA multicycle analysis module, an integrated database, and other utilities. Two benchmark problems were analyzed to confirm the key capabilities of INSIGHT: LP optimization and multicycle analysis. The first was the single cycle LP optimization problem that included various constraints. The second one was the multicycle LP optimization problem that includes the assembly burnup limitation at rod cluster control (RCC) positions. The results for these problems showed the feasibility of INSIGHT for the practical scoping analysis, whose work almost consists of LP generation and multicycle analysis. (author)

  11. Comparison of problems and experience of core operation with distorted fuel element assemblies in VVER-1000 and PWR reactors

    International Nuclear Information System (INIS)

    Afanas'ev, A.

    1999-01-01

    The main reactors leading to distortion of fuel element assemblies during reactor operation were studied. A series of actions which compensate this effect was proposed. Criteria of operation limitation in VVER-1000 and PWR reactors are described

  12. Experiment and analyses on intentional secondary-side depressurization during PWR small break LOCA. Effects of depressurization rate and break area on core liquid level behavior

    International Nuclear Information System (INIS)

    Asaka, Hideaki; Ohtsu, Iwao; Anoda, Yoshinari; Kukita, Yutaka

    1997-01-01

    The effects of the secondary-side depressurization rate and break area on the core liquid level behavior during a PWR small-break LOCA were studied using experimental data from the Large Scale Test Facility (LSTF) and by using analysis results obtained with a JAERI modified version of RELAP5/MOD3 code. The LSTF is a 1/ 48 volumetrically scaled full-height integral model of a Westinghouse-type PWR. The code reproduced the thermal-hydraulic responses, observed in the experiment, for important parameters such as the primary and secondary side pressures and core liquid level behavior. The sensitivity of the core minimum liquid level to the depressurization rate and break area was studied by using the code assessed above. It was found that the core liquid level took a local minimum value for a given break area as a function of secondary side depressurization rate. Further efforts are, however, needed to quantitatively define the maximum core temperature as a function of break area and depressurization rate. (author)

  13. Experimentation, modelling and simulation of water droplets impact on ballooned sheath of PWR core fuel assemblies in a LOCA situation

    International Nuclear Information System (INIS)

    Lelong, Franck

    2010-01-01

    In a pressurized water reactor (PWR), during a Loss Of Coolant Accident (LOCA), liquid water evaporates and the fuel assemblies are not cooled anymore; as a consequence, the temperature rises to such an extent that some parts of the fuel assemblies can be deformed resulting in 'ballooned regions'. When reflooding occurs, the cooling of these partially blocked parts of the fuel assemblies will depend on the coolant flow that is a mixture of overheated vapour and under-saturated droplets. The aim of this thesis is to study the heat transfer between droplets and hot walls of the fuel rods. In this purpose, an experimental device has been designed in accordance with droplets and wall features (droplet velocity and diameter, wall temperature) representative of LOCA conditions. The cooling of a hot Nickel disk, previously heated by induction, is cooled down by a stream of monodispersed droplet. The rear face temperature profiles are measured by infrared thermography. Then, the estimation of wall heat flux is performed by an inverse conduction technique from these infrared images. The effect of droplet dynamical properties (diameter, velocity) on the heat flux is studied. These experimental data allow us to validate an analytical model of heat exchange between droplet and hot slab. This model is based on combined dynamical and thermal considerations. On the one hand, the droplet dynamics is considered through a spring analogy in order to evaluate the evolution of droplet features such as the spreading diameter when the droplet is squeezed over the hot surface. On the other hand, thermal parameters, such as the thickness of the vapour cushion beneath the droplet, are determined from an energy balance. In the short term, this model will be integrated in a CFD code (named NEPTUNE-CFD) to simulate the cooling of a reactor core during a LOCA, taking into account the droplet/wall heat exchange. (author)

  14. An assessment of Class-9 (core-melt) accidents for PWR dry-containment systems

    International Nuclear Information System (INIS)

    Theofanous, T.G.; Saito, M.

    1981-01-01

    The phenomenology of core-melt accidents in dry containments was examined for the purpose of identifying the margins of safety in such Class-9 situations. The scale (geometry) effects appear to crucially limit the extent (severity) of steam explosions. This together with the established reduced explosivity of the corium-A/water system, and the inherently high capability of dry containments (redinforced concrete, and shields in some cases, seismic design etc.) lead to the conclusion that failure due to steam explosions may be considered essentially incredible. These premixture scaling considerations also impact ultimate debris disposition and coolability and need additional development. A water-flooded reactor cavity would have beneficial effects in limiting (but not necessarily eliminating) melt-concrete interactions. Independently of the initial degree of quenching and/or scale of fragmentation, mechanisms exist that drive the system towards ultimate stability (coolability). Additional studies, with intermediate-scale prototypic materials are recommended to better explore these mechanisms. Containment heat removal systems must provide the crucial capability of mitigating such accidents. Passive systems should be explored and assessed against currently available and/or improved active systems taking into account the rather loose time constraints required for activation. It appears that containment margins for accommodating the hydrogen problem are limited. This problem appears to stand out not only in terms of potential consequences but also in terms of lack of any readily available and clear cut solutions at this time. (orig.)

  15. Neutronics characteristics of micro-heterogeneous ThO2-UO2 PWR cores

    International Nuclear Information System (INIS)

    Zhao, X.; Driscoll, M.J.; Kazimi, S.

    2001-01-01

    A new fuel concept, axially-micro-heterogeneous ThO 2 -UO 2 fuel, where ThO 2 fuel pellets and UO 2 fuel pellets are stacked in separate layers in the fuel rods, is being studied at MIT as an option to reduce plutonium production in LWR fuel. Very interesting neutronic behavior is observed: (1) A reactivity increase of 3% to 4% at EOL for a given 235 U inventory which results in a 20-30% increase in average core discharge burnup; (2) For certain configurations, a ''burnable poison'' effect is observed. Analysis shows that these effects are achieved due to a combination of changes in self-shielding, local fissile worth, and conversion ratio, among which self-shielding is the dominant effect at the end of a reactivity-limited burnup. Other variations of micro-heterogeneous UO 2 -ThO 2 fuel including duplex pellets, checkerboard pin distribution, and checkerboard-axial combinations have also been investigated, and their neutronic performance compared. It is concluded that the axial fuel micro-heterogeneity provides the largest gain in reactivity-limited burnup. (author)

  16. Validation of PWR core seismic models with shaking table tests on interacting scale 1 fuel assemblies

    International Nuclear Information System (INIS)

    Viallet, E.; Bolsee, G.; Ladouceur, B.; Goubin, T.; Rigaudeau, J.

    2003-01-01

    The fuel assembly mechanical strength must be justified with respect to the lateral loads under accident conditions, in particular seismic loads. This justification is performed by means of time-history analyses with dynamic models of an assembly row in the core, allowing for assembly deformations, impacts at grid locations and reactor coolant effects. Due to necessary simplifications, the models include 'equivalent' parameters adjusted with respect to dynamic characterisation tests of the fuel assemblies. Complementing such tests on isolated assemblies by an overall model validation with shaking table tests on interacting assemblies is obviously desirable. Seismic tests have been performed by French CEA (Commissariat a l'Energie Atomique) on a row of six full scale fuel assemblies, including two types of 17 x 17 12ft design. The row models are built according to the usual procedure, with preliminary characterisation tests performed on a single assembly. The test-calculation comparisons are made for two test configurations : in air and in water. The relatively large number of accelerograms (15, used for each configuration) is also favourable to significant comparisons. The results are presented for the impact forces at row ends, displacements at mid assembly, and also 'statistical' parameters. Despite a non-negligible scattering in the results obtained with different accelerograms, the calculations prove realistic, and the modelling process is validated with a good confidence level. This satisfactory validation allows to evaluate precisely the margins in the seismic design methodology of the fuel assemblies, and thus to confirm the safety of the plants in case of seismic event. (author)

  17. Performance of core exit thermocouple for PWR accident management action in vessel top break LOCA simulation experiment at OECD/NEA ROSA project

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

    2009-01-01

    Presented are experiment results of the Large Scale Test Facility (LSTF) conducted at the Japan Atomic Energy Agency (JAEA) with a focus on core exit thermocouple (CET) performance to detect core overheat during a vessel top break loss-of-coolant accident (LOCA) simulation experiment. The CET temperatures are used to start accident management (AM) action to quickly depressurize steam generator (SG) secondary side in case of core temperature excursion. Test 6-1 is first test of the OECD/NEA ROSA Project started in 2005, simulating withdraw of a control rod drive mechanism penetration nozzle at the vessel top head. The break size is equivalent to 1.9% cold leg break. The AM action was initiated when CET temperature rose up to 623K. There was no reflux water fallback onto the CETs during the core heat-up period. The core overheat, however, was detected with a time delay of about 230s. In addition, a large temperature discrepancy was observed between the CETs and the hottest core region. This paper clarifies the reason of time delay and temperature discrepancy between the CETs and heated core during boil-off including three-dimensional steam flows in the core and core exit. The paper discusses applicability of the LSTF CET performance to pressurized water reactor (PWR) conditions and a possibility of alternative indicators for earlier AM action than in Test 6-1 is studied by using symptom-based plant parameters such as a reactor vessel water level detection. (author)

  18. Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

    Energy Technology Data Exchange (ETDEWEB)

    Brown, C.S., E-mail: csbrown3@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, 2500 Stinson Drive, Raleigh, NC 27695-7909 (United States); Zhang, H., E-mail: Hongbin.Zhang@inl.gov [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3870 (United States); Kucukboyaci, V., E-mail: kucukbvn@westinghouse.com [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States); Sung, Y., E-mail: sungy@westinghouse.com [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2016-12-01

    Highlights: • Best estimate plus uncertainty (BEPU) analyses of PWR core responses under main steam line break (MSLB) accident. • CASL’s coupled neutron transport/subchannel code VERA-CS. • Wilks’ nonparametric statistical method. • MDNBR 95/95 tolerance limit. - Abstract: VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was applied to simulate core behavior of a typical Westinghouse-designed 4-loop pressurized water reactor (PWR) with 17 × 17 fuel assemblies in response to two main steam line break (MSLB) accident scenarios initiated at hot zero power (HZP) at the end of the first fuel cycle with the most reactive rod cluster control assembly stuck out of the core. The reactor core boundary conditions at the most DNB limiting time step were determined by a system analysis code. The core inlet flow and temperature distributions were obtained from computational fluid dynamics (CFD) simulations. The two MSLB scenarios consisted of the high and low flow situations, where reactor coolant pumps either continue to operate with offsite power or do not continue to operate since offsite power is unavailable. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this demonstration of BEPU application, 59 full core simulations were performed for each accident scenario to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. A parametric goodness-of-fit approach was also applied to the results to obtain the MDNBR value at the 95/95 tolerance limit. Initial sensitivity analysis was performed with the 59 cases per accident scenario by use of Pearson correlation coefficients. The results show that this typical PWR core

  19. Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

    International Nuclear Information System (INIS)

    Brown, C.S.; Zhang, H.; Kucukboyaci, V.; Sung, Y.

    2016-01-01

    Highlights: • Best estimate plus uncertainty (BEPU) analyses of PWR core responses under main steam line break (MSLB) accident. • CASL’s coupled neutron transport/subchannel code VERA-CS. • Wilks’ nonparametric statistical method. • MDNBR 95/95 tolerance limit. - Abstract: VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was applied to simulate core behavior of a typical Westinghouse-designed 4-loop pressurized water reactor (PWR) with 17 × 17 fuel assemblies in response to two main steam line break (MSLB) accident scenarios initiated at hot zero power (HZP) at the end of the first fuel cycle with the most reactive rod cluster control assembly stuck out of the core. The reactor core boundary conditions at the most DNB limiting time step were determined by a system analysis code. The core inlet flow and temperature distributions were obtained from computational fluid dynamics (CFD) simulations. The two MSLB scenarios consisted of the high and low flow situations, where reactor coolant pumps either continue to operate with offsite power or do not continue to operate since offsite power is unavailable. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this demonstration of BEPU application, 59 full core simulations were performed for each accident scenario to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. A parametric goodness-of-fit approach was also applied to the results to obtain the MDNBR value at the 95/95 tolerance limit. Initial sensitivity analysis was performed with the 59 cases per accident scenario by use of Pearson correlation coefficients. The results show that this typical PWR core

  20. 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)

  1. Review of some problems encountered with In-Core Fission chambers and Self-Powered Neutron Detectors in PWR's. Tests - Present use - Outlook on the near future

    International Nuclear Information System (INIS)

    Duchene, Jean; Verdant, Robert.

    1979-01-01

    The working conditions of in-core detectors are investigated as well as some reliability problems which depend on nuclear environment (such as decrease of sensibility, loss of insulation...). Then we review the long-term irradiation tests in experimental reactor that have been carried out by the CEA these last years, with fission chambers (FC) and Self-Powered Detectors (SPD). The travelling probe system with moveable FC used in the 900 MWe PWR is briefly described. Finally an outlook on future possibilities is given; for instance the use of fixed SPD and a moveable FC in the same thimble, allowing recalibration of the fixed detectors [fr

  2. Development of neutron own codes for the simulation of PWR reactor core; Desarrollo de codigos neutronicos propios para la simulacion del nucleo de reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Ahnert, C.; Cabellos, O.; Garcia-Herranz, N.; Cuervo, D.; Herrero, J. J.; Jimenez, J.; Ochoa, R.

    2011-07-01

    The core physic simulation is enough complex to need computers and ad-hoc software, and its evolution is to best-estimate methodologies, in order to improve availability and safety margins in the power plant operation. the Nuclear Engineering Department (UPM) has developed the SEANAP System in use in several power plants in Spain, with simulation in 3D and at the pin level detail, of the nominal and actual core burnup, with the on-line surveillance, and operational maneuvers optimization. (Author) 8 refs.

  3. Application of perturbation theory to sensitivity calculations of PWR type reactor cores using the two-channel model; Aplicacao da teoria de perturbacao para calculos de sensibilidade em nucleos de reatores PWR, usando um modelo de dois canais

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, A.C.J.G. de

    1988-12-01

    Sensitivity calculations are very important in design and safety of nuclear reactor cores. Large codes with a great number of physical considerations have been used to perform sensitivity studies. However, these codes need long computation time involving high costs. The perturbation theory has constituted an efficient and economical method to perform sensitivity analysis. The present work is an application of the perturbation theory (matricial formalism) to a simplified model of DNB (Departure from Nucleate Boiling) analysis to perform sensitivity calculations in PWR cores. Expressions to calculate the sensitivity coefficients of enthalpy and coolant velocity with respect to coolant density and hot channel area were developed from the proposed model. The CASNUR.FOR code to evaluate these sensitivity coefficients was written in Fortran. The comparison between results obtained from the matricial formalism of perturbation theory with those obtained directly from the proposed model makes evident the efficiency and potentiality of this perturbation method for nuclear reactor cores sensitivity calculations (author). 23 refs, 4 figs, 7 tabs.

  4. Performance Specification Shippinpark Pressurized Water Reactor Fuel Drying and Canister Inerting System for PWR Core 2 Blanket Fuel Assemblies Stored within Shippingport Spent Fuel Canisters

    International Nuclear Information System (INIS)

    JOHNSON, D.M.

    2000-01-01

    This specification establishes the performance requirements and basic design requirements imposed on the fuel drying and canister inerting system for Shippingport Pressurized Water Reactor (PWR) Core 2 blanket fuel assemblies (BFAs) stored within Shippingport spent fuel (SSFCs) canisters (fuel drying and canister inerting system). This fuel drying and canister inerting system is a component of the U.S. Department of Energy, Richland Operations Office (RL) Spent Nuclear Fuels Project at the Hanford Site. The fuel drying and canister inerting system provides for removing water and establishing an inert environment for Shippingport PWR Core 2 BFAs stored within SSFCs. A policy established by the U.S. Department of Energy (DOE) states that new SNF facilities (this is interpreted to include structures, systems and components) shall achieve nuclear safety equivalence to comparable U.S. Nuclear Regulatory Commission (NRC)-licensed facilities. This will be accomplished in part by applying appropriate NRC requirements for comparable NRC-licensed facilities to the fuel drying and canister inerting system, in addition to applicable DOE regulations and orders

  5. A Comparative Physics Study of Commercial PWR Cores using Metallic Micro-cell UO{sub 2}-Cr (or Mo) Pellets with Cr-based Cladding Coating

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hee; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of); In, Wang Kee [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this work, a comparative neutronic analysis of the cores using ATFs which include metallic micro-cell UO{sub 2}-Cr, UO{sub 2}-Mo pellets and Cr-based alloy coating on cladding was performed to show the effects of the ATF fuels on the core performance. In this study, the cores having different ATFs use the same initial uranium enrichments. The ATF concepts studied in this work are the metallic microcell UO{sub 2} pellets containing Cr or Mo with cladding outer coating composed of Cr-based alloy which have been suggested as the ATF concepts in KAERI (Korea Atomic Energy Research Institute). The metallic micro-cell pellets and Cr-based alloy coating can enhance thermal conductivity of fuel and reduce the production of hydrogen from the reaction of cladding with coolant, respectively. The objective of this work is to compare neutronic characteristics of commercial PWR equilibrium cores utilizing the different variations of metallic micro-cell UO{sub 2} pellets with cladding coating composed of Cr-based alloy. The results showed that the cores using UO{sub 2}-Cr and UO{sub 2}-Mo pellets with Cr-based alloy coating on cladding have reduced cycle lengths by 60 and 106 EFPDs, respectively, in comparison with the reference UO{sub 2} fueled core due to the reduced heavy metal inventories and large thermal absorption cross section but they do not have any significant differences in the core performances parameters. However, it is notable that the core fueled the micro-cell UO{sub 2}-Mo pellet and Cr-based alloy coating has considerably more negative MTC and slightly more negative FTC than the other cases. These characteristics of the core using micro-cell UO{sub 2}-Mo pellet and Cr-based alloy coating is due to the hard neutron spectrum and large capture resonance cross section of Mo isotopes.

  6. 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)

  7. Verification of NUREC Code Transient Calculation Capability Using OECD NEA/US NRC PWR MOX/UO2 Core Transient Benchmark Problem

    International Nuclear Information System (INIS)

    Joo, Hyung Kook; Noh, Jae Man; Lee, Hyung Chul; Yoo, Jae Woon

    2006-01-01

    In this report, we verified the NUREC code transient calculation capability using OECD NEA/US NRC PWR MOX/UO2 Core Transient Benchmark Problem. The benchmark problem consists of Part 1, a 2-D problem with given T/H conditions, Part 2, a 3-D problem at HFP condition, Part 3, a 3-D problem at HZP condition, and Part 4, a transient state initiated by a control rod ejection at HZP condition in Part 3. In Part 1, the results of NUREC code agreed well with the reference solution obtained from DeCART calculation except for the pin power distributions at the rodded assemblies. In Part 2, the results of NUREC code agreed well with the reference DeCART solutions. In Part 3, some results of NUREC code such as critical boron concentration and core averaged delayed neutron fraction agreed well with the reference PARCS 2G solutions. But the error of the assembly power at the core center was quite large. The pin power errors of NUREC code at the rodded assemblies was much smaller the those of PARCS code. The axial power distribution also agreed well with the reference solution. In Part 4, the results of NUREC code agreed well with those of PARCS 2G code which was taken as the reference solution. From the above results we can conclude that the results of NUREC code for steady states and transient states of the MOX loaded LWR core agree well with those of the other codes

  8. Steady-state and transient core feasibility analysis for a thorium-fuelled reduced-moderation PWR performing full transuranic recycle

    International Nuclear Information System (INIS)

    Lindley, Benjamin A.; Ahmad, Ali; Zainuddin, N. Zara; Franceschini, Fausto; Parks, Geoffrey T.

    2014-01-01

    Highlights: • We present a core analysis for a thorium-transuranic fuelled reduced-moderation PWR. • There is the possibility of positive reactivity in severe large break LOCAs. • Mechanical shim is used to control reactivity within power peaking constraints. • Adequate shutdown margin can be achieved with B 4 C control rods are required. • The response to a rod ejection accident is within likely licensing limits. - Abstract: It is difficult to perform multiple recycle of transuranic (TRU) isotopes in PWRs as the moderator temperature coefficient (MTC) tends to become positive after a few recycles and the core may have positive reactivity when fully voided. Due to the favourable impact on the MTC fostered by use of thorium (Th), the possibility of performing Th–TRU multiple-recycle in reduced-moderation PWRs (RMPWRs) is under consideration. Heterogeneous fuel design with spatial separation of Th–U and Th–TRU is necessary to improve neutronic performance. This can take the form of a heterogeneous fuel assembly (TPUC), or whole assembly heterogeneity (WATU). Satisfactory discharge burn-up can be maintained while ensuring negative MTC, with the pin diameter of a standard PWR increased from 9.5 to 11 mm. However, the reactivity becomes positive when the coolant density in the core becomes extremely low. This could lead to positive reactivity in some loss of coolant accident (LOCA) scenarios, for example a surge line break, if the reactor does not trip. To protect against this beyond design basis accident, a second redundant set of shutdown rods is added to the reactor, so that either the usual or secondary rods can trip the reactor when there is zero coolant in the core. Even so, this condition is likely to be concerning from a regulatory standpoint. Reactivity control is a key challenge due to the reduced worth of neutron absorbers and their detrimental effect on the void coefficients, especially when diluted, as is the case for soluble boron

  9. Model for the probability of core uncovery in loss of offsite power induced accidents, as applied in the Probabilistic Safety Study for ENEL PWR standard power plant

    International Nuclear Information System (INIS)

    Silvestri, E.; Serra, S.; Paddleford, D.F.

    1985-01-01

    This paper discusses one particular aspect of the Probabilistic Safety Study conducted for the Italian reference PWR or Progetto Unificato Nucleare (PUN) design. The event scenario addressed involves the loss of offsite power (LOOSP) initiating event in conjunction with an independent loss of certain support systems (to the exclusion of the total independent loss of on-site power which is treated similarly in a separate event tree). An event tree is developed to address the potential for a consequential small LOCA due to reactor coolant pump (RCP) seal failure under conditions of inadequate seal cooling and the subsequent potential for core uncovery should emergency systems be unavailable and not recovered in adequate time. The event scenario and the quantification methodology used are described. Results and sensitivities are presented

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  11. Comparison of european computer codes relative to the aerosol behavior in PWR containment buildings during severe core damage accidents

    International Nuclear Information System (INIS)

    Fermandjian, J.; Beonio-Brocchieri, F.

    1986-09-01

    The present study concerns a comparative exercise, performed within the framework of the Commission of the European Communities, of the computer codes used in reactor safety in order to assess their capability of realistically describing the aerosol behavior in PWR reactor containment buildings during severe accidents. The codes included in the present study are the following: AEROSIM-M, AEROSOLS/Bl, CORRAL-2, NAUA Mod5. In AEROSIM-M, AEROSOLS/Bl and NAUA Mod5, the integro-differential equation for the evolution of the particle mass distribution is approximated by a set of coupled first order differential equations. To this end, the particle distribution function is replaced by a number of discrete monodisperse fractions. The CORRAL-2 has an essentially empirical basis (processes not explicitely modelled, but their net effects accounted for). The physical processes taken into account in the codes are shown finally

  12. Quasi-Y-source inverter

    DEFF Research Database (Denmark)

    Siwakoti, Yam Prasad; Blaabjerg, Frede; Loh, Poh Chiang

    2015-01-01

    This paper introduces a new inverter topology called a “quasi-Y-source inverter”. The proposed inverter inherits all the advantages of the original Y-source inverter. In addition, the new topology draws continuous current from the source which is required for many renewable sources. It also has dc......-current-blocking capacitors, which avoids saturation in the transformer core. Simulations and experimental results have proved the validity of the proposed inverter....

  13. Determination of the level of water in the core of reactors PWR using neutron detectors signal ex core; Determinacion del nivel del agua del nucleo de reactores PWR usando la senal de detectores neutronicos excore

    Energy Technology Data Exchange (ETDEWEB)

    Bernal, A.; Abarca, A.; Miro, R.; Verdu, G.

    2014-07-01

    The level of water from the core provides relevant information of the neutronic and thermal hydraulic of the reactor as the power, k EFF and cooling capacity. In fact, this level monitoring can be used for prediction of LOCA and reduction of cooling that can cause damage to the core. There are several teams that measure a variety of parameters of the reactor, as opposed to the level of the water of the core. However, the detectors 'excore' measure fast neutrons which escape from the core and there are studies that demonstrate the existence of a relationship between them and the water level of the kernel due to the water shield. Therefore, a methodology has been developed to determine this relationship, using the Monte Carlo method using the MCNP code and apply variance reduction techniques based on the attached flow that is obtained using the method of discrete ordinates using code TORT. (Author)

  14. 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)

  15. 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.)

  16. 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.)

  17. A small long-cycle PWR core design concept using fully ceramic micro-encapsulated (FCM) and UO2–ThO2 fuels for burning of TRU

    International Nuclear Information System (INIS)

    Bae, Gonghoon; Hong, Ser Gi

    2015-01-01

    In this paper, a new small pressurized water reactor (PWR) core design concept using fully ceramic micro-encapsulated (FCM) particle fuels and UO 2 –ThO 2 fuels was studied for effective burning of transuranics from a view point of core neutronics. The core of this concept rate is 100 MWe. The core designs use the current PWR-proven technologies except for a mixed use of the FCM and UO 2 –ThO 2 fuel pins of low-enriched uranium. The significant burning of TRU is achieved with tri-isotropic particle fuels of FCM fuel pins, and the ThO 2 –UO 2 fuel pins are employed to achieve long-cycle length of ∼4 EFPYs (effective full-power year). Also, the effects of several candidate materials for reflector are analyzed in terms of core neutronics because the small core size leads to high sensitivity of reflector material on the cycle length. The final cores having 10 w/o SS303 and 90 w/o graphite reflector are shown to have high TRU burning rates of 33%–35% in FCM pins and significant net burning rates of 24%–25% in the total core with negative reactivity coefficients, low power peaking factors, and sufficient shutdown margins of control rods. (author)

  18. A methodology for evaluating weighting functions using MCNP and its application to PWR ex-core analyses

    International Nuclear Information System (INIS)

    Pecchia, Marco; Vasiliev, Alexander; Ferroukhi, Hakim; Pautz, Andreas

    2017-01-01

    Highlights: • Evaluation of neutron source importance for a given tally. • Assessment of ex-core detector response plus its uncertainty. • Direct use of neutron track evaluated by a Monte Carlo neutron transport code. - Abstract: The ex-core neutron detectors are commonly used to control reactor power in light water reactors. Therefore, it is relevant to understand the importance of a neutron source to the ex-core detectors response. In mathematical terms, this information is conveniently represented by the so called weighting functions. A new methodology based on the MCNP code for evaluating the weighting functions starting from the neutron history database is presented in this work. A simultaneous evaluation of the weighting functions in a user-given Cartesian coverage mesh is the main advantage of the method. The capability to generate weighting functions simultaneously in both spatial and energy ranges is the innovative part of this work. Then, an interpolation tool complements the methodology, allowing the generation of weighting functions up to the pin-by-pin fuel segment, where a direct evaluation is not possible due to low statistical precision. A comparison to reference results provides a verification of the methodology. Finally, an application to investigate the role of ex-core detectors spatial location and core burnup for a Swiss nuclear power plant is provided.

  19. SEDRIO/INCORE, an automatic optimal loading pattern search system for PWR NPP reload core using an expert system

    International Nuclear Information System (INIS)

    Xian Chunyu; Zhang Zongyao

    2003-01-01

    The expert knowledge library for Daya Bay and Qinshan phase II NPP has been established based on expert knowledge, and the reload core loading pattern heuristic search is performed. The in-core fuel management code system INCORE that has been used in engineering design is employed for neutron calculation, and loading pattern is evaluated by using of cycle length and core radial power peaking factor. The developed system SEDRIO/INCORE has been applied in cycle 4 for unit 2 of Daya Bay NPP and cycle 4 for Phase II in Qinshan NPP. The application demonstrated that the loading patterns obtained by SEDRIO/INCORE system are much better than reference ones from the view of the radial power peak and the cycle length

  20. Contribution to the modelling of flows and heat transfers during the reflooding phase of a PWR core

    International Nuclear Information System (INIS)

    Colas, D.

    1984-01-01

    This thesis contributes to modelise thermohydraulic phenomena occuring in a pressurized water nuclear reactor core during the reflood phase of a LOCA. The reference accident and phenomena occuring during reflooding are described as well as flow regime and heat transfer proposed models. With these models, we developed a code to compute fluid conditions and fuel rods temperatures in a reactor core chanel. In order to test this code, results of computation are compared with experiments (FLECHT Skewed Tests) and a conclusion is drawn [fr

  1. Potential of thorium-based fuel cycle for PWR core to reduce plutonium and long-term toxicity

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Hyung Kook; Kim, Taek Kyum; Kim, Young Jin [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-01-01

    The cross section libraries and calculation methods of the participants were inter-compared through the first stage benchmark calculation. The multiplication factor of unit cell benchmark are in good agreement, but there is significant discrepancies of 2.3 to 3.5 %k at BOC and at EOC between the calculated infinite multiplication factors of each participants for the assembly benchmark. Our results with HELIOS show a reasonable agreement with the others except the MTC value at EOC. To verify the potential of the thorium-based fuel to consume the plutonium and to reduce the radioactivity from the spent fuel, the conceptual core with ThO{sub 2}-PuO{sub 2} or MOX fuel were constructed. The composition and quantity of plutonium isotopes and the radioactivity level of spent fuel for conceptual cores were analyzed, and the neutronic characteristics of conceptual cores were also calculated. The nuclear characteristics for ThO{sub 2}-PuO{sub 2} thorium fueled core was similar to MOX fueled core, mainly due to the same seed fuel material, plutonium. For the capability of plutonium consumption, ThO{sub 2}-PuO{sub 2} thorium fuel can consume plutonium 2.1-2.4 times MOX fuel. The fraction of fissile plutonium in the spent ThO{sub 2}-PuO{sub 2} thorium fuel is more favorable in view of plutonium consumption and non-proliferation than MOX fuel. The radioactivity of spent ThO{sub 2}-PuO{sub 2} thorium and MOX fuel batches were calculated. Since plutonium isotopes are dominant for the long-term radioactivity, ThO{sub 2}-PuO{sub 2} thorium has almost the same level of radioactivity as in MOX fuel for a long-term perspective. (author). 22 figs., 11 tabs.

  2. Three dimensions transport calculations for PWR core; Calcul de coeur R.E.P. en transport 3D

    Energy Technology Data Exchange (ETDEWEB)

    Richebois, E

    2000-07-01

    The objective of this work is to define improved 3-D core calculation methods based on the transport theory. These methods can be particularly useful and lead to more precise computations in areas of the core where anisotropy and steep flux gradients occur, especially near interface and boundary conditions and in regions of high heterogeneity (bundle with absorbent rods). In order to apply the transport theory a new method for calculating reflector constants has been developed, since traditional methods were only suited for 2-group diffusion core calculations and could not be extrapolated to transport calculations. In this thesis work, the new method for obtaining reflector constants is derived regardless of the number of energy groups and of the operator used. The core calculations results using the reflector constants thereof obtained have been validated on the EDF's power reactor Saint Laurent B1 with MOX loading. The advantages of a 3-D core transport calculation scheme have been highlighted as opposed to diffusion methods; there are a considerable number of significant effects and potential advantages to be gained in rod worth calculations for instance. These preliminary results obtained with on particular cycle will have to be confirmed by more systematic analysis. Accidents like MSLB (main steam line break) and LOCA (loss of coolant accident) should also be investigated and constitute challenging situations where anisotropy is high and/or flux gradients are steep. This method is now being validated for others EDF's PWRs' reactors, as well as for experimental reactors and other types of commercial reactors. (author)

  3. Three dimensions transport calculations for PWR core; Calcul de coeur R.E.P. en transport 3D

    Energy Technology Data Exchange (ETDEWEB)

    Richebois, E

    2000-07-01

    The objective of this work is to define improved 3-D core calculation methods based on the transport theory. These methods can be particularly useful and lead to more precise computations in areas of the core where anisotropy and steep flux gradients occur, especially near interface and boundary conditions and in regions of high heterogeneity (bundle with absorbent rods). In order to apply the transport theory a new method for calculating reflector constants has been developed, since traditional methods were only suited for 2-group diffusion core calculations and could not be extrapolated to transport calculations. In this thesis work, the new method for obtaining reflector constants is derived regardless of the number of energy groups and of the operator used. The core calculations results using the reflector constants thereof obtained have been validated on the EDF's power reactor Saint Laurent B1 with MOX loading. The advantages of a 3-D core transport calculation scheme have been highlighted as opposed to diffusion methods; there are a considerable number of significant effects and potential advantages to be gained in rod worth calculations for instance. These preliminary results obtained with on particular cycle will have to be confirmed by more systematic analysis. Accidents like MSLB (main steam line break) and LOCA (loss of coolant accident) should also be investigated and constitute challenging situations where anisotropy is high and/or flux gradients are steep. This method is now being validated for others EDF's PWRs' reactors, as well as for experimental reactors and other types of commercial reactors. (author)

  4. PWR in-core nuclear fuel management optimization utilizing nodal (non-linear NEM) generalized perturbation theory

    International Nuclear Information System (INIS)

    Maldonado, G.I.; Turinsky, P.J.; Kropaczek, D.J.

    1993-01-01

    The computational capability of efficiently and accurately evaluate reactor core attributes (i.e., k eff and power distributions as a function of cycle burnup) utilizing a second-order accurate advanced nodal Generalized Perturbation Theory (GPT) model has been developed. The GPT model is derived from the forward non-linear iterative Nodal Expansion Method (NEM) strategy, thereby extending its inherent savings in memory storage and high computational efficiency to also encompass GPT via the preservation of the finite-difference matrix structure. The above development was easily implemented into the existing coarse-mesh finite-difference GPT-based in-core fuel management optimization code FORMOSA-P, thus combining the proven robustness of its adaptive Simulated Annealing (SA) multiple-objective optimization algorithm with a high-fidelity NEM GPT neutronics model to produce a powerful computational tool used to generate families of near-optimum loading patterns for PWRs. (orig.)

  5. Study of the spatial dependence of neutronic flow oscillations caused by fluctuations thermohydraulics at the entrance of the core of a reactor PWR; Estudio de la dependencia espacial de las oscilaciones de flujo neutronico causadas por flucturaciones termohidraulicas a la entrada del nucleo de un reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Bermejo, J. A.; Lopez, A.; Ortego, A.

    2014-07-01

    It presents a theoretical study on spatial dependence of flow oscillations neutronic caused by thermal hydraulics fluctuations at the entrance of the core of a PWR reactor. To simulate, with SIMULATE code - 3K different fluctuations thermohydraulics at the entrance to the core and the spatial dependence of the oscillations and is analyzed neutronic flow obtained at locations of neutron detectors. the work It is part of the r and d program initiated in CNAT to investigate the phenomenon of the noise neutronic. (Author)

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  7. Investigation into fuel pin reshuffling options in PWR in-core fuel management for enhancement of efficient use of nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Daing, Aung Tharn, E-mail: atdaing@khu.ac.kr; Kim, Myung Hyun, E-mail: mhkim@khu.ac.kr

    2014-07-01

    Highlights: • This paper discusses an alternative option, fuel pin reshuffling for maximization of cycle energy production. • The prediction results of isotopic compositions of each burnt pin are verified. • The operating performance is analyzed at equilibrium core with fuel pin reshuffling. • The possibility of reuse of spent fuel pins for reduction of fresh fuel assemblies is investigated. - Abstract: An alternative way to enhance efficient use of nuclear fuel is investigated through fuel pin reshuffling options within PWR fuel assembly (FA). In modeling FA with reshuffled pins, as prerequisite, the single pin calculation method is proposed to estimate the isotopic compositions of each pin of burnt FA in the core-wide environment. Subsequently, such estimation has been verified by comparing with the neutronic performance of the reference design. Two scenarios are concerned, i.e., first scenario was targeted on the improvement of the uniform flux spatial distribution and on the enhancement of neutron economy by simply reshuffling the existing fuel pins in once-burnt fuel assemblies, and second one was focused on reduction of fresh fuel loading and discharged fuel assemblies with more economic incentives by reusing some available spent fuel pins still carrying enough reactivity that are mechanically sound ascertained. In scenario-1, the operating time was merely somewhat increased for few minutes when treating eight FAs by keeping enough safety margins. The scenario-2 was proved to reduce four fresh FAs loading without largely losing any targeted parameters from the safety aspect despite loss of 14 effective full power days for operation at reference plant full rated power.

  8. Droplet generation during core reflood

    International Nuclear Information System (INIS)

    Kocamustafaogullari, G.; De Jarlais, G.; Ishii, M.

    1983-01-01

    The process of entrainment and disintegration of liquid droplets by a flow of steam has considerable practical importance in calculating the effectivenes of the emergency core cooling system. Liquid entrainment is also important in determination of the critical heat flux point in general. Thus the analysis of the reflooding phase of a LOCA requires detailed knowledge of droplet size. Droplet size is mainly determined by the droplet generation mechanisms involved. To study these mechanisms, data generated in the PWR FLECHT SEASET series of experiments was analyzed. In addition, an experiment was performed in which the hydrodynamics of low quality post-CHF flow (inverted annular flow) were simulated in an adiabatic test section

  9. Integrated TRAC/MELPROG analysis of core damage from a severe feedwater transient in the Oconee-1 PWR

    International Nuclear Information System (INIS)

    Henninger, R.J.; Boyack, B.E.

    1986-01-01

    A postulated complete loss-of-feedwater event in the Oconee-1 pressurized water reactor has been analyzed. With an initial version of the lonked TRAC and MELPROG codes, we have modeled the loss-of-feedwater event from initiation to the time of complete disruption of the core, which was calculated to occur by 6800 s. The highest structure temperatures otuside the vessel are on the flow path from the vessel to the pressurizer relief valve. Temperatures in excess of 1200 K could result in failure and depressurization of the primary system before vessel failure

  10. 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

  11. Design and preliminary analysis of in-vessel core catcher made of high-temperature ceramics material in PWR

    International Nuclear Information System (INIS)

    Xu Hong; Ma Li; Wang Junrong; Zhou Zhiwei

    2011-01-01

    In order to protect the interior wall of pressure vessel from melting, as an additional way to external reactor vessel cooling (ERVC), a kind of in-vessel core catcher (IVCC) made of high-temperature ceramics material was designed. Through the high-temperature and thermal-resistance characteristic of IVCC, the distributing of heat flux was optimized. The results show that the downward average heat flux from melt in ceramic layer reduces obviously and the interior wall of pressure vessel doesn't melt, keeping its integrity perfectly. Increasing of upward heat flux from metallic layer makes the upper plenum structure's temperature ascend, but the temperature doesn't exceed its melting point. In conclusion, the results indicate the potential feasibility of IVCC made of high-temperature ceramics material. (authors)

  12. 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)

  13. In-vessel core melt retention by RPV external cooling for high power PWR. MAAP 4 analysis on a LBLOCA scenario without SI

    International Nuclear Information System (INIS)

    Cognet, C.; Gandrille, P.

    1999-01-01

    In-, ex-vessel reflooding or both simultaneously can be envisaged as Accident Management Measures to stop a Severe Accident (SA) in vessel. This paper addresses the possibility of in-vessel core melt retention by RPV external flooding for a high power PWR (4250 MWth). The reactor vessel is assumed to have no lower head penetration and thermal insulation is neglected. The effects of external cooling of high power density debris, where the margin for such a strategy is low, are investigated with the MAAP4 code. MAAP4 code is used to verify the system capability to flood the reactor pit and to predict simultaneously the corium relocation into the lower head with the thermal and mechanical response of the RPV in transient conditions. The corium pool cooling and holding in the RPV lower head is analysed. Attention is paid to the internal heat exchanges between corium components. This paper focuses particularly the heat transfer between oxidic and metallic phases as well as between the molten metallic phase and the RPV wall of utmost importance for challenging the RPV integrity in vicinity of the metallic phase. The metal segregation has a decisive influence upon the attack of the vessel wall due to a very strong peaking of the lateral flux ('focusing effect'). Thus, the dynamics of the formation of the metallic layer characterized by a growing inventory of steel, both from a partial vessel ablation and the degradation of internals steel structures by the radiative heat flux from the debris, is displayed. The analysed sequence is a surge line rupture near the hot leg (LBLOCA) leading to the fastest accident progression

  14. Inverted oscillator

    Energy Technology Data Exchange (ETDEWEB)

    Yuce, C [Physics Department, Anadolu University, Eskisehir (Turkey); Kilic, A [Physics Department, Anadolu University, Eskisehir (Turkey); Coruh, A [Physics Department, Sakarya University, Sakarya (Turkey)

    2006-07-15

    The inverted harmonic oscillator problem is investigated quantum mechanically. The exact wavefunction for the confined inverted oscillator is obtained and it is shown that the associated energy eigenvalues are discrete, and the energy is given as a linear function of the quantum number n.

  15. The effect of metallothionein 2A core promoter region single-nucleotide polymorphism on accumulation of toxic metals in sinonasal inverted papilloma tissues

    Energy Technology Data Exchange (ETDEWEB)

    Starska, Katarzyna, E-mail: katarzyna.starska@umed.lodz.pl [I Department of Otolaryngology and Laryngological Oncology, Medical University of Łódź, Kopcinskiego 22, 90-153 Łódź (Poland); Bryś, Magdalena; Forma, Ewa [Department of Cytobiochemistry, University of Łódź, Pomorska 142/143, 90-236 Łódź (Poland); Olszewski, Jurek; Pietkiewicz, Piotr [II Department of Otolaryngology and Laryngological Oncology, Medical University of Łódź, Żeromskiego 113, 90-549 Łódź (Poland); Lewy-Trenda, Iwona; Danilewicz, Marian [Department of Pathology, Medical University of Łódź, Pomorska 251, 92-213 Łódź (Poland); Krześlak, Anna [Department of Cytobiochemistry, University of Łódź, Pomorska 142/143, 90-236 Łódź (Poland)

    2015-06-15

    Metallothioneins (MTs) are intracellular thiol-rich heavy metal-binding proteins which join trace metal ions protecting cells against heavy metal toxicity and regulate metal distribution and donation to various enzymes and transcription factors. The goal of this study was to identify the − 5 A/G (rs28366003) single-nucleotide polymorphism (SNP) in the core promoter region of the MT2A gene, and to investigate its effect on allele-specific gene expression and Cd, Zn, Cu and Ni content in sinonasal inverted papilloma tissue (IP), with non-cancerous sinonasal mucosa (NCM) as a control. The MT2A promoter region − 5 A/G SNP was identified by restriction fragment length polymorphism using 117 IP and 132 NCM. MT2A gene analysis was performed by quantitative real-time PCR. Metal levels were analyzed by flame atomic absorption spectrometry. The frequency of A allele carriage was 99.2% and 100% in IP and NCM, respectively. The G allele carriage was detected in 23.9% of IP and in 12.1% of the NCM samples. As a result, a significant association of − 5 A/G SNP in MT2A gene with mRNA expression in both groups was determined. A significant association was identified between the − 5 A/G SNP in the MT2A gene with mRNA expression in both groups. A highly significant association was detected between the rs28366003 genotype and Cd and Zn content in IP. Furthermore, significant differences were identified between A/A and A/G genotype with regard to the type of metal contaminant. The Spearman rank correlation results showed the MT2A gene expression and both Cd and Cu levels were negatively correlated. The results obtained in this study suggest that the − 5 A/G SNP in the MT2A gene may have an effect on allele-specific gene expression and toxic metal accumulation in sinonasal inverted papilloma. - Highlights: • MT2A gene expression and metal content in sinonasal inverted papilloma tissues • Association between SNP (rs28366003) and expression of MT2A • Significant

  16. Research of natural resources saving by design studies of Pressurized Light Water Reactors and High Conversion PWR cores with mixed oxide fuels composed of thorium/uranium/plutonium

    International Nuclear Information System (INIS)

    Vallet, V.

    2012-01-01

    Within the framework of innovative neutronic conception of Pressurized Light Water Reactors (PWR) of 3. generation, saving of natural resources is of paramount importance for sustainable nuclear energy production. This study consists in the one hand to design high Conversion Reactors exploiting mixed oxide fuels composed of thorium/uranium/plutonium, and in the other hand, to elaborate multi-recycling strategies of both plutonium and 233 U, in order to maximize natural resources economy. This study has two main objectives: first the design of High Conversion PWR (HCPWR) with mixed oxide fuels composed of thorium/uranium/plutonium, and secondly the setting up of multi-recycling strategies of both plutonium and 233 U, to better natural resources economy. The approach took place in four stages. Two ways of introducing thorium into PWR have been identified: the first is with low moderator to fuel volume ratios (MR) and ThPuO 2 fuel, and the second is with standard or high MR and ThUO 2 fuel. The first way led to the design of under-moderated HCPWR following the criteria of high 233 U production and low plutonium consumption. This second step came up with two specific concepts, from which multi-recycling strategies have been elaborated. The exclusive production and recycling of 233 U inside HCPWR limits the annual economy of natural uranium to approximately 30%. It was brought to light that the strong need in plutonium in the HCPWR dedicated to 233 U production is the limiting factor. That is why it was eventually proposed to study how the production of 233 U within PWR (with standard MR), from 2020. It was shown that the anticipated production of 233 U in dedicated PWR relaxes the constraint on plutonium inventories and favours the transition toward a symbiotic reactor fleet composed of both PWR and HCPWR loaded with thorium fuel. This strategy is more adapted and leads to an annual economy of natural uranium of about 65%. (author) [fr

  17. Application of the Particle Swarm Optimization (PSO) technique to the thermal-hydraulics project of a PWR reactor core in reduced scale

    International Nuclear Information System (INIS)

    Lima Junior, Carlos Alberto de Souza

    2008-09-01

    The reduced scale models design have been employed by engineers from several different industries fields such as offshore, spatial, oil extraction, nuclear industries and others. Reduced scale models are used in experiments because they are economically attractive than its own prototype (real scale) because in many cases they are cheaper than a real scale one and most of time they are also easier to build providing a way to lead the real scale design allowing indirect investigations and analysis to the real scale system (prototype). A reduced scale model (or experiment) must be able to represent all physical phenomena that occurs and further will do in the real scale one under operational conditions, e.g., in this case the reduced scale model is called similar. There are some different methods to design a reduced scale model and from those two are basic: the empiric method based on the expert's skill to determine which physical measures are relevant to the desired model; and the differential equation method that is based on a mathematical description of the prototype (real scale system) to model. Applying a mathematical technique to the differential equation that describes the prototype then highlighting the relevant physical measures so the reduced scale model design problem may be treated as an optimization problem. Many optimization techniques as Genetic Algorithm (GA), for example, have been developed to solve this class of problems and have also been applied to the reduced scale model design problem as well. In this work, Particle Swarm Optimization (PSO) technique is investigated as an alternative optimization tool for such problem. In this investigation a computational approach, based on particle swarm optimization technique (PSO), is used to perform a reduced scale two loop Pressurized Water Reactor (PWR) core, considering 100% of nominal power operation on a forced flow cooling circulation and non-accidental operating conditions. A performance comparison

  18. 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.

  19. 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

  20. 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)

  1. The simulation research for the dynamic performance of integrated PWR

    International Nuclear Information System (INIS)

    Yuan Jiandong; Xia Guoqing; Fu Mingyu

    2005-01-01

    The mathematical model of the reactor core of integrated PWR has been studied and simplified properly. With the lumped parameter method, authors have established the mathematical model of the reactor core, including the neutron dynamic equation, the feedback reactivities model and the thermo-hydraulic model of the reactor. Based on the above equations and models, the incremental transfer functions of the reactor core model have been built. By simulation experimentation, authors have compared the dynamic characteristics of the integrated PWR with the traditional dispersed PWR. The simulation results show that the mathematical models and equations are correct. (authors)

  2. A scaling study of the natural circulation flow of the ex-vessel core catcher cooling system of a 1400MW PWR for designing a scale-down test facility

    International Nuclear Information System (INIS)

    Rhee, Bo. W.; Ha, K. S.; Park, R. J.; Song, J. H.

    2012-01-01

    A scaling study on the steady state natural circulation flow along the flow path of the ex-vessel core catcher cooling system of 1400MWe PWR is described. The scaling criteria for reproducing the same thermalhydraulic characteristics of the natural circulation flow as the prototype core catcher cooling system in the scale-down test facility is derived and the resulting natural circulation flow characteristics of the prototype and scale-down facility analyzed and compared. The purpose of this study is to apply the similarity law to the prototype EU-APR1400 core catcher cooling system and the model test facility of this prototype system and derive a relationship between the heating channel characteristics and the down-comer piping characteristics so as to determine the down-comer pipe size and the orifice size of the model test facility. As the geometry and the heating wall heat flux of the heating channel of the model test facility will be the same as those of the prototype core catcher cooling system except the width of the heating channel is reduced, the axial distribution of the coolant quality (or void fraction) is expected to resemble each other between the prototype and model facility. Thus using this fact, the down-comer piping design characteristics of the model facility can be determined from the relationship derived from the similarity law

  3. 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

  4. 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.

  5. 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

  6. Severe accident considerations for modern KWU-PWR plants

    International Nuclear Information System (INIS)

    Eyink, J.

    1987-01-01

    In assumption of severe accident on modern KWU-PWR plants the author discusses on the: selection of core meltdown sequences, course of the accident, containment behaviour and source terms for fission products release to the environment

  7. 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)

  8. 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

  9. Loss of Coolant Accident (LOCA) / Emergency Core Coolant System (ECCS Evaluation of Risk-Informed Margins Management Strategies for a Representative Pressurized Water Reactor (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    A Risk Informed Safety Margin Characterization (RISMC) toolkit and methodology are proposed for investigating nuclear power plant core, fuels design and safety analysis, including postulated Loss-of-Coolant Accident (LOCA) analysis. This toolkit, under an integrated evaluation model framework, is name LOCA toolkit for the US (LOTUS). This demonstration includes coupled analysis of core design, fuel design, thermal hydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results.

  10. Minor actinide transmutation on PWR burnable poison rods

    International Nuclear Information System (INIS)

    Hu, Wenchao; Liu, Bin; Ouyang, Xiaoping; Tu, Jing; Liu, Fang; Huang, Liming; Fu, Juan; Meng, Haiyan

    2015-01-01

    Highlights: • Key issues associated with MA transmutation are the appropriate loading pattern. • Commercial PWRs are the only choice to transmute MAs in large scale currently. • Considerable amount of MA can be loaded to PWR without disturbing k eff markedly. • Loading MA to PWR burnable poison rods for transmutation is an optimal loading pattern. - Abstract: Minor actinides are the primary contributors to long term radiotoxicity in spent fuel. The majority of commercial reactors in operation in the world are PWRs, so to study the minor actinide transmutation characteristics in the PWRs and ultimately realize the successful minor actinide transmutation in PWRs are crucial problem in the area of the nuclear waste disposal. The key issues associated with the minor actinide transmutation are the appropriate loading patterns when introducing minor actinides to the PWR core. We study two different minor actinide transmutation materials loading patterns on the PWR burnable poison rods, one is to coat a thin layer of minor actinide in the water gap between the zircaloy cladding and the stainless steel which is filled with water, another one is that minor actinides substitute for burnable poison directly within burnable poison rods. Simulation calculation indicates that the two loading patterns can load approximately equivalent to 5–6 PWR annual minor actinide yields without disturbing the PWR k eff markedly. The PWR k eff can return criticality again by slightly reducing the boric acid concentration in the coolant of PWR or removing some burnable poison rods without coating the minor actinide transmutation materials from PWR core. In other words, loading minor actinide transmutation material to PWR does not consume extra neutron, minor actinide just consumes the neutrons which absorbed by the removed control poisons. Both minor actinide loading patterns are technically feasible; most importantly do not need to modify the configuration of the PWR core and

  11. Coupling of 3-D core computational codes and a reactor simulation software for the computation of PWR reactivity accidents induced by thermal-hydraulic transients

    International Nuclear Information System (INIS)

    Raymond, P.; Caruge, D.; Paik, H.J.

    1994-01-01

    The French CEA has recently developed a set of new computer codes for reactor physics computations called the Saphir system which includes CRONOS-2, a three-dimensional neutronic code, FLICA-4, a three-dimensional core thermal hydraulic code, and FLICA-S, a primary loops thermal-hydraulic transient computation code, which are coupled and applied to analyze a severe reactivity accident induced by a thermal hydraulic transient: the Steamline Break accident for a pressurized water reactor until soluble boron begins to accumulate in the core. The coupling of these codes has proved to be numerically stable. 15 figs., 7 refs

  12. Thermal-hydraulic study of integrated steam generator in PWR

    International Nuclear Information System (INIS)

    Osakabe, Masahiro

    1989-01-01

    One of the safety aspects of innovative reactor concepts is the integration of steam generators (SGs) into the reactor vessel in the case of the pressurized water reactor (PWR). All of the reactor system components including the pressurizer are within the reactor vessel in the SG integrated PWR. The simple heat transfer code was developed for the parametric study of the integrated SG. The code was compared to the once-through 19-tube SG experiment and the good agreement between the experimental results and the code predictions was obtained. The assessed code was used for the parametric study of the integrated once-through 16 m-straight-tube SG installed in the annular downcomer. The proposed integrated SG as a first attempt has approximately the same tube size and pitch as the present PWR and the SG primary and secondary sides in the present PWR is inverted in the integrated PWR. Based on the study, the reactor vessel size of the SG integrated PWR was calculated. (author)

  13. Study of the functionalization of cross sections of cell in multi groups for calculations in best-estimate 3D pin-by-pin of PWR cores

    International Nuclear Information System (INIS)

    Hueso, C.; Sanchez-Cervera, S.; Herrero, J. J.

    2011-01-01

    One of the objectives of the European project NURISP (Nuclear Reactor Integrated Platform) of 7th framework Programme is to advance the simulation of light water reactors by coupling the best-estimate codes to deepen core physics, thermal-hydraulic behaviour of biphasic and fuel.

  14. Determination of the hydrogen source term during the reflooding of an overheated core: Calculation results of the integral reflood test QUENCH-03 with PWR-type bundle

    International Nuclear Information System (INIS)

    Chikhi, Nourdine; Nguyen, Nam Giang; Fleurot, Joelle

    2012-01-01

    Highlights: ► Calculation of QUENCH-03 experiment with ASTEC/CATHARE. ► Validation of reflooding model in severe accidents conditions. ► Demonstration of a minimum flow rate for a successful reflood by using a system code. ► Effect of injection flow rate on hydrogen production. ► Effect of initial core temperature on hydrogen production. - Abstract: During a severe accident, one of the main accident management procedure consists of injecting water in the reactor core by means of various safety injection devices. Nevertheless, the success of a core reflood is not guaranteed because of possible negative effects: temperature escalation, enhanced hydrogen production, enhanced release of fission products, core degradation due to thermal shock, shattering, debris and melt formation. The QUENCH-03 experiment was carried out to investigate the behavior on reflooding at high temperature of LWR fuel rods with little oxidation. Posttest calculations with the ASTEC-CATHARE V2 code were made for code assessment and validation of the new reflooding model. This thermal–hydraulic model is used to detect the quench front position and to calculate the heat transfer between fuel and fluid in the transition boiling region. Comparisons between the calculational and experimental results are presented. Emphasis has been placed on clad temperature, hydrogen production and melt relocation. The effects of core state damage (initial temperature at reflooding onset) and the reflood mass flow rate on the hydrogen source term were investigated using the QUENCH-03 test as a base case. Calculations were made by varying both parameters in the input data deck. The results demonstrate (and confirm) the existence of a minimum flow rate for a successful reflood.

  15. Inverted annular flow experimental study

    International Nuclear Information System (INIS)

    De Jarlais, G.; Ishii, M.

    1985-04-01

    Steady-state inverted annular flow of Freon 113 in up flow was established in a transparent test section. Using a special inlet configuration consisting of long aspect-ratio liquid nozzles coaxially centered within a heated quartz tube, idealized inverted annular flow initial geometry (cylindrical liquid core surrounded by coaxial annulus of gas) could be established. Inlet liquid and gas flowrates, liquid subcooling, and gas density (using various gas species) were measured and varied systematically. The hydrodynamic behavior of the liquid core, and the subsequent downstream break-up of this core into slugs, ligaments and/or droplets of various sizes, was observed. In general, for low inlet liquid velocities it was observed that after the initial formation of roll waves on the liquid core surface, an agitated region of high surface area, with attendant high momentum and energy transfers, occurs. This agitated region appears to propagate downsteam in a quasi-periodic pattern. Increased inlet liquid flow rates, and high gas annulus flow rates tend to diminish the significance of this agitated region. Observed inverted annular flow (and subsequent downstream flow pattern) hydrodynamic behavior is reported, and comparisons are drawn to data generated by previous experimenters studying post-CHF flow

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

  17. 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)

  18. Behaviour of a PWR with core protection system (SSN) in case of accidents due to power failure, ATWS and steam generator rupture

    International Nuclear Information System (INIS)

    Boncompagni, S.; Fulceri, P.; Oriolo, F.

    1985-01-01

    The results of the analysis of the transient fallowing internal and external power failure, without scram, in the nuclear power plant of the Italian Unified Nuclear Project are examined. The availability of ECCS is excluded while the breakage of a tube in each steam generator is supposed, togheter with the presence of an original safety system known as SSN (core protection system). Computations have been performed by using Mark 6 RELAP4 code. The study of the transient and the physical model used are briefly illustrated. Finally the results achieved are analysed

  19. Sensitivity of risk parameters to human errors for a PWR

    International Nuclear Information System (INIS)

    Samanta, P.; Hall, R.E.; Kerr, W.

    1980-01-01

    Sensitivities of the risk parameters, emergency safety system unavailabilities, accident sequence probabilities, release category probabilities and core melt probability were investigated for changes in the human error rates within the general methodological framework of the Reactor Safety Study for a Pressurized Water Reactor (PWR). Impact of individual human errors were assessed both in terms of their structural importance to core melt and reliability importance on core melt probability. The Human Error Sensitivity Assessment of a PWR (HESAP) computer code was written for the purpose of this study

  20. In-Core Fuel Managements for PWRs: Investigation on solution for optimal utilization of PWR fuel through the use of fuel assemblies with differently enriched 235U fuel pins

    International Nuclear Information System (INIS)

    Caprioli, Sara

    2004-04-01

    A possibility for more efficient use of the nuclear fuel in a pressurized water reactor is investigated. The alternative proposed here consists of the implementation of PWR fuel assemblies with differently enriched 235 U fuel pins. This possibility is examined in comparison with the standard assembly design. The comparison is performed both in terms of single assembly performance and in the terms of nuclear reactor core performance and fuel utility. For the evaluation of the actual performance of the new assembly types, 5 operated fuel core sequences of R3 (Ringhals' third unit), for the period 1999 - 2004 (cycles 17 - 21) were examined. For every cycle, the standard fresh fuel assemblies have been identified and taken as reference cases for the study of the new type of assemblies with differently enriched uranium rods. In every cycle, assemblies with and without burnable absorber are freshly loaded into the core. The axial enrichment distribution is kept uniform, allowing for a radial (planar) enrichment level distribution only. At an assembly level, it has been observed that the implementation of the alternative enrichment configuration can lead to lower and flatter internal peaking factor distribution with respect to the uniformly enriched reference assemblies. This can be achieved by limiting the enrichment levels distribution to a rather narrow range. The highest enrichment level chosen has the greatest impact on the power distribution of the assemblies. As it increases, the enrichment level drives the internal peaking factor to greater values than in the reference assemblies. Generally, the highest enrichment level that would allow an improvement in the power performance of the assembly lies between 3.95 w/o and 4.17 w/o. The highest possible enrichment level depends on the average enrichment of the overall assembly, which is kept constant to the average enrichment of the reference assemblies. The improvements that can be obtained at this level are rather

  1. The research on burnup characteristic of doping burnable poison in PWR

    International Nuclear Information System (INIS)

    Qiang Shenglong; Qin Dong; Chai Xiaoming; Yao Dong

    2014-01-01

    In PWR core design, burnable poisons are usually used for reactive compensation and power flatten. The choice of burnable poisons and how to match burnup would be the key-points for a long-life core design. We study the burnup character of doping burnable poisons (such as natural element, manual nuclide and soluble boron) in the PWR by the core burnup code MOI based on Monte Carlo method. The results show that Hf, Er and Eu doping burnable poison would be applicable for the nuclear design research on the long-life PWR core. (authors)

  2. Status of developing advanced PWR in Japan

    International Nuclear Information System (INIS)

    Iida, Yotaro

    1982-01-01

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

  3. Evaluation of the radiative transfer in the core of a Pressurized Water Reactor (PWR) during the reflooding step of a Loss Of Coolant Accident (LOCA)

    International Nuclear Information System (INIS)

    Gerardin, J.

    2012-01-01

    We developed a method of resolution of radiative transfer inside a medium of vapor-droplets surrounded by hot walls, in order to couple it with a simulation of the flow at the CFD scale. The scope is the study of the cooling of the core of nuclear reactor following a Loss Of Coolant Accident (LOCA). The problem of radiative transfer can be cut into two sub problems, one concerning the evaluation of the radiative properties of the medium and a second concerning the solution of the radiative transfer equation. The radiative properties of the droplets have been computed with the use of the Mie Theory and those of the vapor have been computed with a Ck model. The medium made of vapor and droplets is an absorbing, anisotropically scattering, emissive, non grey, non homogeneous medium. Hence, owing to the possible variations of the flow properties (diameter and volumetric fraction of the droplets, temperature and pressure of the vapor), the medium can be optically thin or thick. Consequently, a method is required which solves the radiative transfer accurately, with a moderate calculation time for all of these prerequisites. The IDA has been chosen, derived from the well-known P1-approximation. Its accuracy has been checked on academical cases found in the literature and by comparison with experimental data. Simulations of LOCA flows have been conducted taking account of the radiative transfer, evaluating the radiative fluxes and showing that radiative transfer influence cannot be neglected. (author)

  4. In-Core Fuel Managements for PWRs: Investigation on solution for optimal utilization of PWR fuel through the use of fuel assemblies with differently enriched {sup 235}U fuel pins

    Energy Technology Data Exchange (ETDEWEB)

    Caprioli, Sara

    2004-04-01

    A possibility for more efficient use of the nuclear fuel in a pressurized water reactor is investigated. The alternative proposed here consists of the implementation of PWR fuel assemblies with differently enriched {sup 235}U fuel pins. This possibility is examined in comparison with the standard assembly design. The comparison is performed both in terms of single assembly performance and in the terms of nuclear reactor core performance and fuel utility. For the evaluation of the actual performance of the new assembly types, 5 operated fuel core sequences of R3 (Ringhals' third unit), for the period 1999 - 2004 (cycles 17 - 21) were examined. For every cycle, the standard fresh fuel assemblies have been identified and taken as reference cases for the study of the new type of assemblies with differently enriched uranium rods. In every cycle, assemblies with and without burnable absorber are freshly loaded into the core. The axial enrichment distribution is kept uniform, allowing for a radial (planar) enrichment level distribution only. At an assembly level, it has been observed that the implementation of the alternative enrichment configuration can lead to lower and flatter internal peaking factor distribution with respect to the uniformly enriched reference assemblies. This can be achieved by limiting the enrichment levels distribution to a rather narrow range. The highest enrichment level chosen has the greatest impact on the power distribution of the assemblies. As it increases, the enrichment level drives the internal peaking factor to greater values than in the reference assemblies. Generally, the highest enrichment level that would allow an improvement in the power performance of the assembly lies between 3.95 w/o and 4.17 w/o. The highest possible enrichment level depends on the average enrichment of the overall assembly, which is kept constant to the average enrichment of the reference assemblies. The improvements that can be obtained at this level are

  5. PWR and WWER fuel performance. A comparison of major characteristics

    International Nuclear Information System (INIS)

    Weidinger, H.

    2006-01-01

    PWR and WWER fuel technologies have the same basic performance targets: most effective use of the energy stored in the fuel and highest possible reliability. Both fuel technologies use basically the same strategies to reach these targets: 1) Optimized reload strategies; 2) Maximal use of structural material with low neutron cross sections; 3) Decrease the fuel failure frequency towards a 'zero failure' performance by understanding and eliminating the root causes of those defects. The key driving force of the technology of both, PWR and WWER fuel is high burn-up. Presently a range of 45 - 50 MWD/kgU have been reached commercially for PWR and WWER fuel. The main technical limitations to reach high burn-up are typically different for PWR and WWER fuel: for PWR fuel it is the corrosion and hydrogen uptake of the Zr-based materials; for WWER fuel it is the mechanical and dimensional stability of the FA (and the whole core). Corrosion and hydrogen uptake of Zr-materials is a 'non-problem' for WWER fuel. Other performance criteria that are important for high burn-up are the creep and growth behaviour of the Zr materials and the fission gas release in the fuel rod. There exists a good and broad data base to model and design both fuel types. FA and fuel rod vibration appears to be a generic problem for both fuel types but with more evidence for PWR fuel performance reliability. Grid-to-rod fretting is still a major issue in the fuel failure statistics of PWR fuel. Fuel rod cladding defects by debris fretting is no longer a key problem for PWR fuel, while it still appears to be a significant root cause for WWER fuel failures. 'Zero defect' fuel performance is achievable with a high probability, as statistics for US PWR and WWER-1000 fuel has shown

  6. 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)

  7. Transient performance of flow in circuits of PWR type reactors

    International Nuclear Information System (INIS)

    Hirdes, V.R.; Carajilescov, P.

    1988-09-01

    Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which could cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

  8. Transient performance of flow in PWR reactor circuits

    International Nuclear Information System (INIS)

    Hirdes, V.R.T.R.; Carajilescov, P.

    1988-12-01

    Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

  9. Application of the Particle Swarm Optimization (PSO) technique to the thermal-hydraulics project of a PWR reactor core in reduced scale; Aplicacao da tecnica de otimizacao por enxame de particulas no projeto termo-hidraulico em escala reduzida do nucleo de um reator PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lima Junior, Carlos Alberto de Souza

    2008-09-15

    The reduced scale models design have been employed by engineers from several different industries fields such as offshore, spatial, oil extraction, nuclear industries and others. Reduced scale models are used in experiments because they are economically attractive than its own prototype (real scale) because in many cases they are cheaper than a real scale one and most of time they are also easier to build providing a way to lead the real scale design allowing indirect investigations and analysis to the real scale system (prototype). A reduced scale model (or experiment) must be able to represent all physical phenomena that occurs and further will do in the real scale one under operational conditions, e.g., in this case the reduced scale model is called similar. There are some different methods to design a reduced scale model and from those two are basic: the empiric method based on the expert's skill to determine which physical measures are relevant to the desired model; and the differential equation method that is based on a mathematical description of the prototype (real scale system) to model. Applying a mathematical technique to the differential equation that describes the prototype then highlighting the relevant physical measures so the reduced scale model design problem may be treated as an optimization problem. Many optimization techniques as Genetic Algorithm (GA), for example, have been developed to solve this class of problems and have also been applied to the reduced scale model design problem as well. In this work, Particle Swarm Optimization (PSO) technique is investigated as an alternative optimization tool for such problem. In this investigation a computational approach, based on particle swarm optimization technique (PSO), is used to perform a reduced scale two loop Pressurized Water Reactor (PWR) core, considering 100% of nominal power operation on a forced flow cooling circulation and non-accidental operating conditions. A performance

  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. Y-source inverter

    DEFF Research Database (Denmark)

    Siwakoti, Yam P.; Town, Graham; Loh, Poh Chiang

    2014-01-01

    This paper introduces a new 3-phase Y-source inverter whose gain is presently not matched by classical impedance-network-based inverters operating at the same duty ratio. The proposed network uses a tightly coupled transformer with three windings. By squeezing the shoot-through range while keeping...... higher boost, the inverter can operate at a higher modulation index, thereby minimizing switching device stress and providing better output power quality. In addition, the inverter has more degrees of freedom for setting the voltage gain and modulation index than other classical impedance-source networks...

  12. Live Imaging of Shoot Meristems on an Inverted Confocal Microscope Using an Objective Lens Inverter Attachment

    Science.gov (United States)

    Nimchuk, Zachary L.; Perdue, Tony D.

    2017-01-01

    Live imaging of above ground meristems can lead to new insights in plant development not possible from static imaging of fixed tissue. The use of an upright confocal microscope offers several technical and biological advantages for live imaging floral or shoot meristems. However, many departments and core facilities possess only inverted confocal microscopes and lack the funding for an additional upright confocal microscope. Here we show that imaging of living apical meristems can be performed on existing inverted confocal microscopes with the use of an affordable and detachable InverterScope accessory. PMID:28579995

  13. Live Imaging of Shoot Meristems on an Inverted Confocal Microscope Using an Objective Lens Inverter Attachment.

    Science.gov (United States)

    Nimchuk, Zachary L; Perdue, Tony D

    2017-01-01

    Live imaging of above ground meristems can lead to new insights in plant development not possible from static imaging of fixed tissue. The use of an upright confocal microscope offers several technical and biological advantages for live imaging floral or shoot meristems. However, many departments and core facilities possess only inverted confocal microscopes and lack the funding for an additional upright confocal microscope. Here we show that imaging of living apical meristems can be performed on existing inverted confocal microscopes with the use of an affordable and detachable InverterScope accessory.

  14. Full MOX high burn-up PWR

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, Tsutomu; Kugo, Teruhiko; Shimada, Shoichiro; Araya, Fumimasa; Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    As a part of conceptual investigation on advanced light water reactors for the future, a light water reactor with the high burn-up of 100 GWd/t, the long cycle operation of 3 years and the full MOX core is being studied, aiming at the improvement on economical aspects, the reduction of the spent fuel production, the utilization of Plutonium and so forth. The present report summarizes investigation on PWR-type reactors. The core with the increased moderation of the moderator-to-fuel volume ratio of 2.6 {approx} 3.0 has been proposed be such a core that accomplishes requirements mentioned above. Through the neutronic and the thermo-hydrodynamic evaluation, the performances of the core have been evaluated. Also, the safety designing is underway considering the reactor system with the passive safety features. (author)

  15. Study on thermo-hydraulic behavior during reflood phase of a PWR-LOCA

    International Nuclear Information System (INIS)

    Sugimoto, Jun

    1989-01-01

    This paper describes thermo-hydraulic behavior during the reflood phase in a postulated large-break loss-of-coolant accident (LOCA) of a PWR. In order to better predict the reflood transient in a nuclear safety analysis specific analytical models have been developed for, saturated film boiling heat transfer in inverted slung flow, the effect of grid spacers on core thermo-hydraulics, overall system thermo-hydraulic behavior, and the thermal response similarity between nuclear fuel rods and simulated rods. A heat transfer correlation has been newly developed for saturated film boiling based on a 4 x 4-rod experiment conducted at JAERI. The correlation provides a good agreement with existing experiments except in the vicinity of grid spacer locations. An analytical model has then been developed addressing the effect of grid spacers. The thermo-hydraulic behavior near the grid spacers was found to be predicted well with this model by considering the breakup of droplets in dispersed flow and water accumulation above the grid spacers in inverted slung flow. A system analysis code has been developed which couples the one-dimensional core and multi-loop primary system component models. It provides fairly good agreement with system behavior obtained in a large-scale integral reflood experiment with active primary system components. An analytical model for the radial temperature distribution in a rod has been developed and verified with data from existing experiments. It was found that a nuclear fuel rod has a lower cladding temperature and an earlier quench time than an electrically heated rod in a typical reflood condition. (author)

  16. 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)

  17. 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.)

  18. Next Generation Inverter

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Zilai [General Motors LLC, Detroit, MI (United States); Gough, Charles [General Motors LLC, Detroit, MI (United States)

    2016-04-22

    The goal of this Cooperative Agreement was the development of a Next Generation Inverter for General Motors’ electrified vehicles, including battery electric vehicles, range extended electric vehicles, plug-in hybrid electric vehicles and hybrid electric vehicles. The inverter is a critical electronics component that converts battery power (DC) to and from the electric power for the motor (AC).

  19. 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

  20. Modular inverter system

    Science.gov (United States)

    Ma, Mingyao; Hu, Haibing; Kutkut, Nasser; Batarseh, Issa; Shen, John; , Bkayrat, Raed

    2017-08-01

    A system connected to an AC power grid having an AC phase signal includes an inverter module including a first inverter coupled to a DC voltage, actuated based on the AC phase signal. The first inverter provides a first voltage signal having predetermined harmonic components. A second inverter includes second switch elements coupled to the DC voltage and actuated by a second set of control signals phase delayed with respect to the first control signals. A transformer module has first and second primary windings coupled to the first and second inverters. The transformer module further includes a secondary winding coupled to first primary winding, the second primary winding, and the AC power grid. The secondary winding is configured to provide a secondary output voltage to the AC power grid by combining the first voltage signal and the second voltage signal such that the predetermined harmonic components are substantially cancelled.

  1. PWR plant operator training used full scope simulator incorporated MAAP model

    International Nuclear Information System (INIS)

    Matsumoto, Y.; Tabuchi, T.; Yamashita, T.; Komatsu, Y.; Tsubouchi, K.; Banka, T.; Mochizuki, T.; Nishimura, K.; Iizuka, H.

    2015-01-01

    NTC makes an effort with the understanding of plant behavior of core damage accident as part of our advanced training. For the Fukushima Daiichi Nuclear Power Station accident, we introduced the MAAP model into PWR operator training full scope simulator and also made the Severe Accident Visual Display unit. From 2014, we will introduce new training program for a core damage accident with PWR operator training full scope simulator incorporated the MAAP model and the Severe Accident Visual Display unit. (author)

  2. Characterization of Decommissioned PWR Vessel Internals Material Samples: Tensile and SSRT Testing (Nonproprietary Version)

    International Nuclear Information System (INIS)

    Krug, M.; Shogan, R.

    2004-01-01

    Pressurized water reactor (PWR) cores operate under extreme environmental conditions due to coolant chemistry, operating temperature, and neutron exposure. Extending the life of PWRs requires detailed knowledge of the changes in mechanical and corrosion properties of the structural austenitic stainless steel components adjacent to the fuel (internals) subjected to such conditions. This project studied the effects of reactor service on the mechanical and corrosion properties of samples of baffle plate, former plate, and core barrel from a decommissioned PWR

  3. Influence of boron reduction strategies on PWR accident management flexibility

    International Nuclear Information System (INIS)

    Papukchiev, Angel Aleksandrov; Liu, Yubo; Schaefer, Anselm

    2007-01-01

    In conventional pressurized water reactor (PWR) designs, soluble boron is used for reactivity control over core fuel cycle. Design changes to reduce boron concentration in the reactor coolant are of general interest regarding three aspects - improved reactivity feedback properties, lower impact of boron dilution scenarios on PWR safety and eventually more flexible accident management procedures. In order to assess the potential advantages through the introduction of boron reduction strategies in current PWRs, two low boron core configurations based on fuel with increased utilization of gadolinium and erbium burnable absorbers have been developed. The new PWR designs permit to reduce the natural boron concentration in reactor coolant at begin of cycle to 518 ppm and 805 ppm. For the assessment of the potential safety advantages of these cores a hypothetical beyond design basis accident has been simulated with the system code ATHLET. The analyses showed improved inherent safety and increased accident management flexibility of the low boron cores in comparison with the standard PWR. (author)

  4. Stochastic optimization of loading pattern for PWR

    International Nuclear Information System (INIS)

    Smuc, T.; Pevec, D.

    1994-01-01

    The application of stochastic optimization methods in solving in-core fuel management problems is restrained by the need for a large number of proposed solutions loading patterns, if a high quality final solution is wanted. Proposed loading patterns have to be evaluated by core neutronics simulator, which can impose unrealistic computer time requirements. A new loading pattern optimization code Monte Carlo Loading Pattern Search has been developed by coupling the simulated annealing optimization algorithm with a fast one-and-a-half dimensional core depletion simulator. The structure of the optimization method provides more efficient performance and allows the user to empty precious experience in the search process, thus reducing the search space size. Hereinafter, we discuss the characteristics of the method and illustrate them on the results obtained by solving the PWR reload problem. (authors). 7 refs., 1 tab., 1 fig

  5. Minimization of PWR reactor control rods wear

    International Nuclear Information System (INIS)

    Ponzoni Filho, Pedro; Moura Angelkorte, Gunther de

    1995-01-01

    The Rod Cluster Control Assemblies (RCCA's) of Pressurized Water Reactors (PWR's) have experienced a continuously wall cladding wear when Reactor Coolant Pumps (RCP's) are running. Fretting wear is a result of vibrational contact between RCCA rodlets and the guide cards which provide lateral support for the rodlets when RCCA's are withdrawn from the core. A procedure is developed to minimize the rodlets wear, by the shuffling and axial reposition of RCCA's every operating cycle. These shuffling and repositions are based on measurement of the rodlet cladding thickness of all RCCA's. (author). 3 refs, 2 figs, 2 tabs

  6. 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

  7. 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.

  8. 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

  9. Monte Carlo based radial shield design of typical PWR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.

    2016-11-15

    Neutron and gamma flux and dose equivalent rate distribution are analysed in radial and shields of a typical PWR type reactor based on the Monte Carlo radiation transport computer code MCNP5. The ENDF/B-VI continuous energy cross-section library has been employed for the criticality and shielding analysis. The computed results are in good agreement with the reference results (maximum difference is less than 56 %). It implies that MCNP5 a good tool for accurate prediction of neutron and gamma flux and dose rates in radial shield around the core of PWR type reactors.

  10. 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

  11. A Multi-Physics PWR Model for the Load Following

    OpenAIRE

    Muniglia , Mathieu; Do , Jean-Michel; Jean-Charles , Le Pallec; Grard , Hubert; Verel , Sébastien; David , S.

    2016-01-01

    International audience; In this paper, a new model of a Pressurized Water Reactor (PWR) is described. This model includes the description of the core as well as a simplified secondary loop: the goal is to reproduce a load-following type transient, where the output power of the plant is controlled by the electric grid. Consequently, the control systems are also modeled, as the control rods or the soluble boron. The reference power plant is a 1300MW electrical PWR, managed with the french G mode.

  12. Multilevel DC link inverter

    Science.gov (United States)

    Su, Gui-Jia

    2003-06-10

    A multilevel DC link inverter and method for improving torque response and current regulation in permanent magnet motors and switched reluctance motors having a low inductance includes a plurality of voltage controlled cells connected in series for applying a resulting dc voltage comprised of one or more incremental dc voltages. The cells are provided with switches for increasing the resulting applied dc voltage as speed and back EMF increase, while limiting the voltage that is applied to the commutation switches to perform PWM or dc voltage stepping functions, so as to limit current ripple in the stator windings below an acceptable level, typically 5%. Several embodiments are disclosed including inverters using IGBT's, inverters using thyristors. All of the inverters are operable in both motoring and regenerating modes.

  13. Reliable inverter systems

    Science.gov (United States)

    Nagano, S.

    1979-01-01

    Base driver with common-load-current feedback protects paralleled inverter systems from open or short circuits. Circuit eliminates total system oscillation that can occur in conventional inverters because of open circuit in primary transformer winding. Common feedback signal produced by functioning modules forces operating frequency of failed module to coincide with clock drive so module resumes normal operating frequency in spite of open circuit.

  14. Study of the noise propagation in PWR with coupled codes

    International Nuclear Information System (INIS)

    Verdu, G.; Garcia-Fenoll, M.; Abarca, A.; Miro, R.; Barrachina, T.

    2011-01-01

    The in-core detectors provide signals of the power distribution monitoring for the Reactor Protection System (RPS). The advanced fuel management strategies (high exposure) and the power upratings for PWR reactor types have led to an increase in the noise amplitude in detectors signals. In the present work a study of the propagation along the reactor core and the effects on the core power evolution of a small perturbation on the moderator density, using the coupled code RELAP5-MOD3.3/PARCSv2.7 is presented. The purpose of these studies is to be able to reproduce and analyze the in-core detector simulated signals. (author)

  15. 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

  16. Hydrodynamic stability of inverted annular flow in an adiabatic simulation

    International Nuclear Information System (INIS)

    De Jarlais, G.; Ishii, M.; Linehan, J.

    1986-01-01

    Inverted annular flow was simulated adiabatically with turbulent water jets, issuing downward from large aspect ratio nozzles, enclosed in gas annuli. Velocities, diameters, and gas species were varied, and core jet length, shape, breakup mode, and dispersed core droplet sizes were recorded at approximately 750 data points. Inverted annular flow destabilization led to inverted slug flow at low relative velocities, and to dispersed droplet flow, core breakup length correlations were developed by extending work on free liquid jets to include this coaxial, jet disintegration phenomenon. The results show length dependence upon D/sub J/, Re/sub J/, We/sub J/, α, and We/sub G/,rel. Correlations for core shape, breakup mechanisms, and dispersed core droplet size were also developed, by extending the results of free jet stability, roll wave entrainment, and churn turbulent droplet stability studies

  17. 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

  18. 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

  19. Simplified model of a PWR primary circuit

    International Nuclear Information System (INIS)

    Souza, A.L.; Faya, A.J.G.

    1988-07-01

    The computer program RENUR was developed to perform a very simplified simulation of a typical PWR primary circuit. The program has mathematical models for the thermal-hydraulics of the reactor core and the pressurizer, the rest of the circuit being treated as a single volume. Heat conduction in the fuel rod is analyzed by a nodal model. Average and hot channels are treated so that bulk response of the core and DNBR can be evaluated. A homogenenous model is employed in the pressurizer. Results are presented for a steady-state situation as well as for a loss of load transient. Agreement with the results of more elaborate computer codes is good with substantial reduction in computer costs. (author) [pt

  20. Resonant snubber inverter

    Science.gov (United States)

    Lai, Jih-Sheng; Young, Sr., Robert W.; Chen, Daoshen; Scudiere, Matthew B.; Ott, Jr., George W.; White, Clifford P.; McKeever, John W.

    1997-01-01

    A resonant, snubber-based, soft switching, inverter circuit achieves lossless switching during dc-to-ac power conversion and power conditioning with minimum component count and size. Current is supplied to the resonant snubber branches solely by the main inverter switches. Component count and size are reduced by use of a single semiconductor switch in the resonant snubber branches. Component count is also reduced by maximizing the use of stray capacitances of the main switches as parallel resonant capacitors. Resonance charging and discharging of the parallel capacitances allows lossless, zero voltage switching. In one embodiment, circuit component size and count are minimized while achieving lossless, zero voltage switching within a three-phase inverter.

  1. Parameterized representation of macroscopic cross section for PWR reactor

    International Nuclear Information System (INIS)

    Fiel, João Cláudio Batista; Carvalho da Silva, Fernando; Senra Martinez, Aquilino; Leal, Luiz C.

    2015-01-01

    Highlights: • This work describes a parameterized representation of the homogenized macroscopic cross section for PWR reactor. • Parameterization enables a quick determination of problem-dependent cross-sections to be used in few group calculations. • This work allows generating group cross-section data to perform PWR core calculations without computer code calculations. - Abstract: The purpose of this work is to describe, by means of Chebyshev polynomials, a parameterized representation of the homogenized macroscopic cross section for PWR fuel element as a function of soluble boron concentration, moderator temperature, fuel temperature, moderator density and 235 92 U enrichment. The cross-section data analyzed are fission, scattering, total, transport, absorption and capture. The parameterization enables a quick and easy determination of problem-dependent cross-sections to be used in few group calculations. The methodology presented in this paper will allow generation of group cross-section data from stored polynomials to perform PWR core calculations without the need to generate them based on computer code calculations using standard steps. The results obtained by the proposed methodology when compared with results from the SCALE code calculations show very good agreement

  2. SEEPAGE/INVERT INTERACTIONS

    International Nuclear Information System (INIS)

    P.S. Domski

    2000-01-01

    As directed by a written development plan (CRWMS M andO 1999a), a conceptual model for water entering the drift and reacting with the invert materials is to be developed. The purpose of this conceptual model is to assist Performance Assessment Operations (PAO) and its Engineered Barrier Performance Department in modeling the geochemical environment within a repository drift, thus allowing PAO to provide a more detailed and complete in-drift geochemical model abstraction, and to answer the key technical issues (KTI) raised in the NRC Issue Resolution Status Report (IRSR) for the Evolution of the Near-Field Environment (NFE), Revision 2 (NRC 1999). This AMR also seeks to: (1) Develop a logical conceptual model for physical/chemical interactions between seepage and the invert materials; (2) screen potential processes and reactions that may occur between seepage and invert to evaluate the potential consequences of the interactions; and (3) outline how seepage/invert processes may be quantified. This document provides the conceptual framework for screening out insignificant processes and for identifying and evaluating those seepage/invert interactions that have the potential to be important to subsequent PAO analyses including the Engineered Barrier System (EBS) physical and chemical model abstraction effort. This model has been developed to serve as a basis for the in-drift geochemical analyses performed by PAO. Additionally, the concepts discussed within this report may also apply to certain near and far-field geochemical processes and may have conceptual application within the unsaturated zone (UZ) and saturated zone (SZ) transport modeling efforts. The seepage/invert interactions will not directly affect any principal factors

  3. Study on thermal-hydraulics during a PWR reflood phase

    Energy Technology Data Exchange (ETDEWEB)

    Iguchi, Tadashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-10-01

    In-core thermal-hydraulics during a PWR reflood phase following a large-break LOCA are quite unique in comparison with two-phase flow which has been studied widely in previous researches, because the geometry of the flow path is complicated (bundle geometry) and water is at extremely low superficial velocity and almost under stagnant condition. Hence, some phenomena realized during a PWR reflood phase are not understood enough and appropriate analytical models have not been developed, although they are important in a viewpoint of reactor safety evaluation. Therefore, author investigated some phenomena specified as important issues for quantitative prediction, i.e. (1) void fraction in a bundle during a PWR reflood phase, (2) effect of radial core power profile on reflood behavior, (3) effect of combined emergency core coolant injection on reflood behavior, and (4) the core separation into two thermal-hydraulically different regions and the in-core flow circulation behavior observed during a combined injection PWR reflood phase. Further, author made analytical models for these specified issues, and succeeded to predict reflood behaviors at representative types of PWRs, i.e.cold leg injection PWRs and Combined injection PWRs, in good accuracy. Above results were incorporated into REFLA code which is developed at JAERI, and they improved accuracy in prediction and enlarged applicability of the code. In the present study, models were intended to be utilized in a practical use, and hence these models are simplified ones. However, physical understanding on the specified issues in the present study is basic and principal for reflood behavior, and then it is considered to be used in a future advanced code development and improvement. (author). 110 refs.

  4. Identification and evaluation of PWR in-vessel severe accident management strategies

    International Nuclear Information System (INIS)

    Dukelow, J.S.; Harrison, D.G.; Morgenstern, M.

    1992-03-01

    This reports documents work performed the NRC/RES Accident Management Guidance Program to evaluate possible strategies for mitigating the consequences of PWR severe accidents. The selection and evaluation of strategies was limited to the in-vessel phase of the severe accident, i.e., after the initiation of core degradation and prior to RPV failure. A parallel project at BNL has been considering strategies applicable to the ex-vessel phase of PWR severe accidents

  5. Characterization of Decommissioned PWR Vessel Internals Materials Samples: Material Certification, Fluence, and Temperature (Nonproprietary Version)

    International Nuclear Information System (INIS)

    Krug, M.; Shogan, R.; Fero, A.; Snyder, M.

    2004-01-01

    Pressurized water reactor (PWR) cores, operate under extreme environmental conditions due to coolant chemistry, operating temperature, and neutron exposure. Extending the life of PWRs require detailed knowledge of the changes in mechanical and corrosion properties of the structural austenitic stainless steel components adjacent to the fuel. This report contains basic material characterization information of the as-installed samples of reactor internals material which were harvested from a decommissioned PWR

  6. Challenging Ubiquitous Inverted Files

    NARCIS (Netherlands)

    de Vries, A.P.

    2000-01-01

    Stand-alone ranking systems based on highly optimized inverted file structures are generally considered ‘the’ solution for building search engines. Observing various developments in software and hardware, we argue however that IR research faces a complex engineering problem in the quest for more

  7. Energy-Saving Inverter

    Science.gov (United States)

    Rippel, W. E.; Edwards, D. B.

    1984-01-01

    Commutation by field-effect transistor allows more efficient operation. High voltage field-effect transistor (FET) controls silicon controlled rectifiers (SCR's). Circuit requires only one capacitor and one inductor in commutation circuit: simpler, more efficient, and more economical than conventional inverters. Adaptable to dc-to-dc converters.

  8. Argonne inverted sputter source

    International Nuclear Information System (INIS)

    Yntema, J.L.; Billquist, P.J.

    1983-01-01

    The emittance of the inverted sputter source with immersion lenses was measured to be about 5π mm mrad MeV/sup 1/2/ at the 75% level over a wide range of beam intensities. The use of the source in experiments with radioactive sputter targets and hydrogen loaded targets is described. Self contamination of the source is discussed

  9. Experimental study of inverted-annular-flow hydrodynamics utilizing an adiabatic simulation

    International Nuclear Information System (INIS)

    De Jarlais, G.

    1983-03-01

    In experiments, inverted annular flow was simulated adiabatically with turbulent water jets, issuing downward from long aspect nozzles, enclosed in gas annuli. Velocities, diameters, and gas species were varied, and core jet length, shape, break-up mode, and dispersed-core droplet sizes were recorded at approximately 750 data points. Inverted annular flow was observed to develop into inverted slug flow at low relative velocities, and into dispersed droplet flow at high relative velocities. For both of the above transitions from inverted annular flow, correlations for core jet length were developed by extending work done on free liquid jets to include this new, coaxial, jet disintegration phenomenon. Jet break-up length is correlated as a function of jet diameter, jet Reynolds number, jet Weber number, void fraction, and gas Weber number. Correlations for core shape, break-up mechanisms and dispersed core droplet size for the case of transition to inverted slug flow were developed

  10. 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

  11. Scope and procedures of fuel management for PWR nuclear power plant

    International Nuclear Information System (INIS)

    Yao Zenghua

    1997-01-01

    The fuel management scope of PWR nuclear power plant includes nuclear fuel purchase and spent fuel disposal, ex-core fuel management, in-core fuel management, core management and fuel assembly behavior follow up. A suit of complete and efficient fuel management procedures have to be created to ensure the quality and efficiency of fuel management work. The hierarchy of fuel management procedure is divided into four levels: main procedure, administration procedure, implement procedure and technic procedure. A brief introduction to the fuel management scope and procedures of PWR nuclear power plant are given

  12. The optimum fuel and power distribution for a PWR burnup cycle

    International Nuclear Information System (INIS)

    Stillman, J.A.

    1989-01-01

    A method was developed to determine the optimum fuel and power distributions for a PWR burnup cycle. The backward diffusion calculation [1] and the Core-wise Green's Function [2] method were used for the core model which provided analytic derivatives for solving the nonlinear optimization problem using successive linear programming [3] methods. The solution algorithm consisted of a reverse depletion strategy which begins at the end of cycle and solves simultaneously for the optimal fuel and burnable absorber distributions while the core is depleted to the beginning of cycle. The resulting optimal solutions minimize the required fissile fuel inventory and burnable absorber loading for a PWR

  13. Inverter communications using output signal

    Science.gov (United States)

    Chapman, Patrick L.

    2017-02-07

    Technologies for communicating information from an inverter configured for the conversion of direct current (DC) power generated from an alternative source to alternating current (AC) power are disclosed. The technologies include determining information to be transmitted from the inverter over a power line cable connected to the inverter and controlling the operation of an output converter of the inverter as a function of the information to be transmitted to cause the output converter to generate an output waveform having the information modulated thereon.

  14. Material property changes of stainless steels under PWR irradiation

    International Nuclear Information System (INIS)

    Fukuya, Koji; Nishioka, Hiromasa; Fujii, Katsuhiko; Kamaya, Masayuki; Miura, Terumitsu; Torimaru, Tadahiko

    2009-01-01

    Structural integrity of core structural materials is one of the key issues for long and safe operation of pressurized water reactors. The stainless steel components are exposed to neutron irradiation and high-temperature water, which cause significant property changes and irradiation assisted stress corrosion cracking (IASCC) in some cases. Understanding of irradiation induced material property changes is essential to predict integrity of core components. In the present study, microstructure and microchemistry, mechanical properties, and IASCC behavior were examined in 316 stainless steels irradiated to 1 - 73 dpa in a PWR. Dose-dependent changes of dislocation loops and cavities, grain boundary segregation, tensile properties and fracture mode, deformation behavior, and their interrelation were discussed. Tensile properties and deformation behavior were well coincident with microstructural changes. IASCC susceptibility under slow strain rate tensile tests, IASCC initiation under constant load tests in simulated PWR primary water, and their relationship to material changes were discussed. (author)

  15. Industrial assessment of nonbackfittable PWR design modifications. Final report

    International Nuclear Information System (INIS)

    Matzie, R.A.; Daleas, R.S.; Miller, D.D.

    1980-11-01

    As part of the US Department of Energy's Advanced Reactor Design Study, various nonbackfittable PWR design modifications were evaluated to determine their potential for improved uranium utilization and commercial viability. Combustion Engineering, Inc. contributed to this effort through participation in the Battelle Pacific Northwest Laboratory industrial assessment of such design modifications. Seven modifications, including the use of higher primary system temperatures and pressures, rapid-frequent refueling, end-of-cycle stretchout, core periphery modifications, radial blankets, low power density cores, and small PWR assemblies, were evaluated with respect to uranium utilization, economics, technical and operational complexity, and several other subjective considerations. Rapid-frequent refueling was judged to have the highest potential although it would probably not be economical for the majority of reactors with the design assumptions used in this assessment

  16. Inverting the Linear Algebra Classroom

    Science.gov (United States)

    Talbert, Robert

    2014-01-01

    The inverted classroom is a course design model in which students' initial contact with new information takes place outside of class meetings, and students spend class time on high-level sense-making activities. The inverted classroom model is so called because it inverts or "flips" the usual classroom design where typically class…

  17. Milliwatt dc/dc Inverter

    Science.gov (United States)

    Mclyman, C. W.

    1983-01-01

    Compact dc/dc inverter uses single integrated-circuit package containing six inverter gates that generate and amplify 100-kHz square-wave switching signal. Square-wave switching inverts 10-volt local power to isolated voltage at another desired level. Relatively high operating frequency reduces size of filter capacitors required, resulting in small package unit.

  18. A new approach to PWR power control using intelligent techniques

    International Nuclear Information System (INIS)

    Boroushaki, M.; Ghofrani, M.B.; Lucas, C.; Yazdanpanah, M.J.; Sadati, N.

    2004-01-01

    Improved load following capability is one of the main technical performances of advanced PWR(APWR). Controlling the nuclear reactor core during load following operation encounters some difficulties. These difficulties mainly arise from nuclear reactor core limitations in local power peaking, while the core is subject to large and sharp variation of local power density during transients. Axial offset (A.O) is the parameter usually used to represent of core power peaking, in form of a practical parameter. This paper, proposes a new intelligent approach to A.o control of PWR nuclear reactors core during load following operation. This method uses a neural network model of the core to predict the dynamic behavior of the core and a fuzzy critic based on the operator knowledge and experience for the purpose of decision-making during load following operations. Simulation results show that this method can use optimum control rod groups maneuver with variable overlapping and may improve the reactor load following capability

  19. Monte Carlo based radial shield design of typical PWR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.

    2017-04-15

    This paper presents the radiation shielding model of a typical PWR (CNPP-II) at Chashma, Pakistan. The model was developed using Monte Carlo N Particle code [2], equipped with ENDF/B-VI continuous energy cross section libraries. This model was applied to calculate the neutron and gamma flux and dose rates in the radial direction at core mid plane. The simulated results were compared with the reference results of Shanghai Nuclear Engineering Research and Design Institute (SNERDI).

  20. Ciclon: A neutronic fuel management program for PWR's consecutive cycles

    International Nuclear Information System (INIS)

    Aragones, J.M.

    1977-01-01

    The program description and user's manual of a new computer code is given. Ciclon performs the neutronic calculation of consecutive reload cycles for PWR's fuel management optimization. Fuel characteristics and burnup data, region or batch sizes, loading schemes and state of previously irradiated fuel are input to the code. Cycle lengths or feed enrichments and burnup sharing for each region or batch are calculate using different core neutronic models and printed or punched in standard fuel management format. (author) [es

  1. SACHET, Dynamic Fission Products Inventory in PWR Multiple Compartment System

    International Nuclear Information System (INIS)

    Kodaira, Hideki

    1990-01-01

    1 - Description of program or function: SACHET evaluates the dynamic fission product inventories in the multiple compartment system of pressurized water reactor (PWR) plants. 2 - Method of solution: SACHET utilizes a matrix of fission product core inventory which is previously calculated by the ORIGEN code. 3 - Restrictions on the complexity of the problem: Liquid wastes such as chemical waste and detergent waste are not included

  2. Irradiation behavior of German PWR RPV steels under operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    May, J.; Hein, H. [AREVA NP Gmbh (Germany); Ganswind, J. [VGB PowerTech e.V. (Germany); Widera, M. [RWE Power AG (Germany)

    2011-07-01

    In 2007, the last standard surveillance capsule of the original RPV (Reactor Pressure Vessel) surveillance programs of the 11 currently operating German PWR has been evaluated. With it the standard irradiation surveillance programs of these plants was completed. In the present paper, irradiation data of these surveillance programs will be presented and a final assessment of the irradiation behavior of the German PWR RPV steels with respect to current standards KTA 3203 and Reg. Guide 1.99 Rev. 2 will be given. Data from two units which are currently under decommissioning will also be included, so that data from all 13 German PWR manufactured by the former Siemens/KWU company (now AREVA NP GmbH) are shown. It will be shown that all surveillance data within the approved area of chemical composition verify the limit curve RT(limit) of the KTA 3203, which is the relevant safety standard for these plants. An analysis of the data shows, that the prediction formulas of Reg. Guide 1.99 Rev. 2 Pos. 1 or from the TTS model tend to overestimate the irradiation behavior of the German PWR RPV steels. Possible reasons for this behavior are discussed. Additionally, the data will be compared to data from the research project CARISMA to demonstrate that these data are representative for the irradiation behavior of the German PWR RPV steels. Since the data of these research projects cover a larger neutron fluence range than the original surveillance data, they offer a future outlook into the irradiation behavior of the German PWR RPV steels under long term conditions. In general, as a consequence of the relatively large and beneficial water gap between core and RPV, especially in all Siemens/KWU 4-loop PWR, the EOL neutron fluence and therefore the irradiation induced changes in mechanical properties of the German PWR RPV materials are rather low. Moreover the irradiation data indicate that the optimized RPV materials specifications that have been applied in particular for the

  3. FLUOLE-2: An Experiment for PWR Pressure Vessel Surveillance

    Directory of Open Access Journals (Sweden)

    Thiollay Nicolas

    2016-01-01

    Full Text Available FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10−2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006–2007 in a geometry representative of 1300 MWe PWR.

  4. Inverter ratio failure detector

    Science.gov (United States)

    Wagner, A. P.; Ebersole, T. J.; Andrews, R. E. (Inventor)

    1974-01-01

    A failure detector which detects the failure of a dc to ac inverter is disclosed. The inverter under failureless conditions is characterized by a known linear relationship of its input and output voltages and by a known linear relationship of its input and output currents. The detector includes circuitry which is responsive to the detector's input and output voltages and which provides a failure-indicating signal only when the monitored output voltage is less by a selected factor, than the expected output voltage for the monitored input voltage, based on the known voltages' relationship. Similarly, the detector includes circuitry which is responsive to the input and output currents and provides a failure-indicating signal only when the input current exceeds by a selected factor the expected input current for the monitored output current based on the known currents' relationship.

  5. PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

    Directory of Open Access Journals (Sweden)

    Virpi Kouhia

    2012-01-01

    Full Text Available This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

  6. PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

    International Nuclear Information System (INIS)

    Virpi Kouhia, V.; Purhonen, H.; Riikonen, V.; Puustinen, M.; Kyrki-Rajamaki, R.; Vihavainen, J.

    2012-01-01

    This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

  7. Toward full MOX core design

    International Nuclear Information System (INIS)

    Rouviere, G.; Guillet, J.L.; Bruna, G.B.; Pelet, J.

    1999-01-01

    This paper presents a selection of the main preliminary results of a study program sponsored by COGEMA and currently carried out by FRAMATOME. The objective of this study is to investigate the feasibility of full MOX core loading in a French 1300 MWe PWR, a recent and widespread standard nuclear power plant. The investigation includes core nuclear design, thermal hydraulic and systems aspects. (authors)

  8. 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.)

  9. PWR type reactor plant

    International Nuclear Information System (INIS)

    Matsuoka, Tsuyoshi.

    1993-01-01

    A water chamber of a horizontal U-shaped pipe type steam generator is partitioned to an upper high temperature water chamber portion and a lower low temperature water chamber portion. An exit nozzle of a reactor container containing a reactor core therein is connected to a suction port of a coolant pump by way of first high temperature pipelines. The exit port of the coolant pump is connected to the high temperature water chamber portion of the steam generator by way of second high temperature pipelines. The low temperature water chamber portion of the steam generator is connected to an inlet nozzle of the reactor container by way of the low temperature pipelines. The low temperature water chamber portion of the steam generator is positioned lower than the high temperature water chamber portion, but upper than the reactor core. Accordingly, all of the steam generator for a primary coolant system, coolant pumps as well as high temperature pipelines and low temperature pipelines connecting them are disposed above the reactor core. With such a constitution, there is no worry of interrupting core cooling even upon occurrence of an accident, to improve plant safety. (I.N.)

  10. Road-map design for thorium-uranium breeding recycle in PWR - 031

    International Nuclear Information System (INIS)

    Shengyi, Si

    2010-01-01

    The paper was focused on designing a road-map to finally approach sustainable Thorium-Uranium ( 232 Th- 233 U) Breeding Recycle in current PWR, without any other change to the fuel lattice and the core internals, but substituting the UOX pellet with Thorium-based pellet. At first, the paper presented some insights to the inherence of Thorium-Uranium fuel conversion or breeding in PWR based on the neutronics theory and revealed the prerequisites for Thorium-Uranium fuel in PWR to achieve sustainable Breeding Recycle; And then, various Thorium-based fuels were designed and examined, and the calculation results further validated the above theoretical deductions; Based on the above theoretical analysis and calculation results, a road-map for sustainable Thorium-Uranium breeding recycle in PWR was outlined finally. (authors)

  11. 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.)

  12. Experimental study on reflooding in advanced tight lattice PWR

    International Nuclear Information System (INIS)

    Hori, K.; Kodama, J.; Teramae, T.

    2000-01-01

    This paper is related to the experimental study on the feasibility of core cooling by re-flooding in a large break loss of coolant accident (LOCA) for the advanced tight lattice pressurized water reactor (PWR). The tight lattice core design should be adopted to improve the conversion ratio. Major one of the key questions of such tight lattice core is the cooling capability under the re-flood condition in a large break LOCA. Forced feed bottom re-flooding experiments have been performed by use of a 4x4 triangular array rod bundle. The rod gap is 0.5 mm, 1.0 mm, or 1.5 mm. The measured peak temperature is below around 1273 K even in case of 1.0/0.5 mm rod gap. And, the evaluation based on the experimental results of rod temperatures and core pressure drop also shows that the core cooling under re-flooding condition is feasible. (author)

  13. 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)

  14. Upper internals of PWR with coolant flow separator

    International Nuclear Information System (INIS)

    Chevereau, G.; Heuze, A.

    1989-01-01

    The upper internals for a PWR has a collecting volume for the coolant merging from the core and an apparatus for separating the flow of coolant. This apparatus has a guide for the control rods, a lower plate perforated to allow the coolant through from the core, an upper plate also perforated to allow the coolant through to the collecting volume and a peripheral binding ring joining the two plates. Each guide comprises an envelope without holes and joined perceptibly tight to the plates [fr

  15. Hydrodynamics of adiabatic inverted annular flow: an experimental study

    International Nuclear Information System (INIS)

    De Jarlais, G.; Ishii, M.

    1983-01-01

    For low-quality film boiling in tubes or rod bundles, the flow pattern may consist of a liquid jet-like core surrounded by a vapor annulus, i.e., inverted annular flow. The stability, shape, and break-up mechanisms of this liquid core must be understood in order to model correctly this regime and to develop appropriate interfacial transfer correlations. This paper reports on a study in which inverted annular flow was simulated in an adiabatic system. Turbulent water jets, issuing downward from long-aspect nozzles were enclosed within cocurrent gas annuli. Jet-core diameter and velocity, and gas-annulus diameter, velocity, and species were varied, yielding liquid Reynolds numbers up to 33,000, void fractions from 0.29 to 0.95, and relative velocities from near zero to over 80 m/s. Jet-core break-up lengths and secondarily, core break-up mechanisms, were observed visually, using strobe lighting

  16. Advances in PV Inverters

    DEFF Research Database (Denmark)

    Anthon, Alexander

    and preferably low complexity leads to new research demands. This is especially true in the field of low cost residential PV inverters where efficiencies are used as major selling arguments. Traditional converter topologies equipped with conventional Silicon based semiconductors to date reach their limitations......, a replacement of only two switching devices per phase leg can greatly reduce the semiconductor losses. The Hybrid-NPC converter can be seen as an attractive and cost competitive alternative to the Silicon Carbide based converter, also allowing to overcome the major drawbacks with the conventional Silicon IGBT...

  17. Mathematical modelling of plant transients in the PWR for simulator purposes

    International Nuclear Information System (INIS)

    Hartel, K.

    1984-01-01

    This chapter presents the results of the testing of anticipated and abnormal plant transients in pressurized water reactors (PWRs) of the type WWER 440 by means of the numerical simulation of 32 different, stationary and nonstationary, operational regimes. Topics considered include the formation of the PWR mathematical model, the physical approximation of the reactor core, the structure of the reactor core model, a mathematical approximation of the reactor model, the selection of numerical methods, and a computerized simulation system. The necessity of a PWR simulator in Czechoslovakia is justified by the present status and the outlook for the further development of the Czechoslovak nuclear power complex

  18. Identification of mechanical vibrations in a PWR reactor using neutron noise signal analysis of the standard instrumentation; Identifikacija mehanichkih varijacija analizom signala shuma standardne neutronske instrumentacije PWR reaktora

    Energy Technology Data Exchange (ETDEWEB)

    Kostic, Lj [Institut za Nuklearne Nauke Boris Kidric, Belgrade (Yugoslavia); Runkel, J [Institut fuer Kerntechnik und Zerstoerungsfreie Pruefverfahren, Hannover (Germany)

    1988-07-01

    The neutron noise signals in a PWR power plant were analysed in terms of auto- and cross-power spectral densities, phases and coherences. Core barrel motion, fuel element vibrations and reactivity noise effect due to pressure variations have been monitored and analysed. (author)

  19. Solar Charged Stand Alone Inverter

    OpenAIRE

    M.Vasugi; Prof R.Jayaraman

    2014-01-01

    This paper deals with solar powered stand alone inverter which converts the variable dc output of a photovoltaic solar panel into ac that can be fed to loads. Stand alone inverters are used in systems where the inverter get its energy from batteries charged by photo voltaic arrays. A charge controller limits the rate at which electric current is added to or drawn from electric batteries. This charge discharge controller is needed to prevent the battery from being overcharged o...

  20. Gas cooled traction drive inverter

    Science.gov (United States)

    Chinthavali, Madhu Sudhan

    2013-10-08

    The present invention provides a modular circuit card configuration for distributing heat among a plurality of circuit cards. Each circuit card includes a housing adapted to dissipate heat in response to gas flow over the housing. In one aspect, a gas-cooled inverter includes a plurality of inverter circuit cards, and a plurality of circuit card housings, each of which encloses one of the plurality of inverter cards.

  1. Polynomial parameterized representation of macroscopic cross section for PWR reactor

    International Nuclear Information System (INIS)

    Fiel, Joao Claudio B.

    2015-01-01

    The purpose of this work is to describe, by means of Tchebychev polynomial, a parameterized representation of the homogenized macroscopic cross section for PWR fuel element as a function of soluble boron concentration, moderator temperature, fuel temperature, moderator density and 235 U 92 enrichment. Analyzed cross sections are: fission, scattering, total, transport, absorption and capture. This parameterization enables a quick and easy determination of the problem-dependent cross-sections to be used in few groups calculations. The methodology presented here will enable to provide cross-sections values to perform PWR core calculations without the need to generate them based on computer code calculations using standard steps. The results obtained by parameterized cross-sections functions, when compared with the cross-section generated by SCALE code calculations, or when compared with K inf , generated by MCNPX code calculations, show a difference of less than 0.7 percent. (author)

  2. PWR plant transient analyses using TRAC-PF1

    International Nuclear Information System (INIS)

    Ireland, J.R.; Boyack, B.E.

    1984-01-01

    This paper describes some of the pressurized water reactor (PWR) transient analyses performed at Los Alamos for the US Nuclear Regulatory Commission using the Transient Reactor Analysis Code (TRAC-PF1). Many of the transient analyses performed directly address current PWR safety issues. Included in this paper are examples of two safety issues addressed by TRAC-PF1. These examples are pressurized thermal shock (PTS) and feed-and-bleed cooling for Oconee-1. The calculations performed were plant specific in that details of both the primary and secondary sides were modeled in addition to models of the plant integrated control systems. The results of these analyses show that for these two transients, the reactor cores remained covered and cooled at all times posing no real threat to the reactor system nor to the public

  3. Serious accidents of PWR type reactors for power generation

    International Nuclear Information System (INIS)

    2008-12-01

    This document presents the great lines of current knowledge on serious accidents relative to PWR type reactors. First, is exposed the physics of PWR type reactor core meltdown and the possible failure modes of the containment building in such a case. Then, are presented the dispositions implemented with regards to such accidents in France, particularly the pragmatic approach that prevails for the already built reactors. Then, the document tackles the case of the European pressurized reactor (E.P.R.), for which the dimensioning takes into account explicitly serious accidents: it is a question of objectives conception and their respect must be the object of a strict demonstration, by taking into account uncertainties. (N.C.)

  4. The plutonium recycle for PWR reactors from brazilian nuclear program

    International Nuclear Information System (INIS)

    Rubini, L.A.

    1978-01-01

    The purpose of this thesis is to evaluate the material requirements of the nuclear fuel cycle with plutonium recycle. The study starts with the calculation of a reference reactor and has flexibility to evaluate the demand under two alternatives of nuclear fuel cycle for Pressurized Water Reactors (PWR): Without plutonium recycle; and with plutonium recycle. Calculations of the reference reactor have been carried out with the CELL-CORE codes. Variations in the material requirements were studied considering changes in the installed nuclear capacity of PWR reactors, the capacity factor of these reactors, and the introduction of fast breeders. Recycling plutonium produced inside the system can reach economies of about 5% U 3 O 8 and 6% separative work units if recycle is assumed only after the fifth operation cycle of the thermal reactors. (author)

  5. Gadolinia experience and design for PWR fuel cycles

    International Nuclear Information System (INIS)

    Stephenson, L. C.

    2000-01-01

    The purpose of this paper is to describe Siemens Power Corporation's (SPC) current experience with the burnable absorber gadolinia in PWR fuel assemblies, including optimized features of SPC's PWR gadolinia designs, and comparisons with other burnable absorbers. Siemens is the world leader in PWR gadolinia experience. More than 5,900 Siemens PWR gadolinia-bearing fuel assemblies have been irradiated. The use of gadolinia-bearing fuel provides significant flexibility in fuel cycle designs, allows for low radial leakage fuel management and extended operating cycles, and reduces BOC (beginning-of-cycle) soluble boron concentrations. The optimized use of an integral burnable neutron absorber is a design feature which provides improved economic performance for PWR fuel assemblies. This paper includes a comparison between three different types of integral burnable absorbers: gadolinia, Zirconium diboride and erbia. Fuel cycle design studies performed by Siemens have shown that the enrichment requirements for 18-24 month fuel cycles utilizing gadolinia or zirconium diboride integral fuel burnable absorbers can be approximately the same. Although a typical gadolinia residual penalty for a cycle design of this length is as low as 0.02-0.03 wt% U-235, the design flexibility of gadolinia allows for very aggressive low-leakage core loading plans which reduces the enrichment requirements for gadolinia-bearing fuel. SPC has optimized its use of gadolinia in PWR fuel cycles. Typically, low (2-4) weight percent Gd 2 O 3 is used for beginning to middle of cycle reactivity hold down as well as soluble boron concentration holddown at BOC. Higher concentrations of Gd 2 O 3 , such as 6 and 8 wt%, are used to control power peaking in assemblies later in the cycle. SPC has developed core strategies that maximize the use of lower gadolinia concentrations which significantly reduces the gadolinia residual reactivity penalty. This optimization includes minimizing the number of rods with

  6. 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

  7. 2015 Inverter Workshop | Photovoltaic Research | NREL

    Science.gov (United States)

    Inverter Workshop 2015 Inverter Workshop Wednesday, February 25, 2015 Chair: Jack Flicker In about inverters. This workshop represented a follow-on to the inverter workshops that Sandia National conversations between module and inverter experts. Agenda For a detailed schedule of the day's events, access

  8. Evaluation of CRUDTRAN code to predict transport of corrosion products and radioactivity in the PWR primary coolant system

    International Nuclear Information System (INIS)

    Lee, C.B.

    2002-01-01

    CRUDTRAN code is to predict transport of the corrosion products and their radio-activated nuclides such as cobalt-58 and cobalt-60 in the PWR primary coolant system. In CRUDTRAN code the PWR primary circuit is divided into three principal sections such as the core, the coolant and the steam generator. The main driving force for corrosion product transport in the PWR primary coolant comes from coolant temperature change throughout the system and a subsequent change in corrosion product solubility. As the coolant temperature changes around the PWR primary circuit, saturation status of the corrosion products in the coolant also changes such that under-saturation in steam generator and super-saturation in the core. CRUDTRAN code was evaluated by comparison with the results of the in-reactor loop tests simulating the PWR primary coolant system and PWR plant data. It showed that CRUDTRAN could predict variations of cobalt-58 and cobalt-60 radioactivity with time, plant cycle and coolant chemistry in the PWR plant. (author)

  9. IVA2 - a computer code for modelling of transient 3D-three phase three component flows using three velocity fields in cylindrical geometry with arbitrary internals including nuclear reactor PWR/BWR-core

    International Nuclear Information System (INIS)

    Kolev, N.I.

    1986-06-01

    This report contains a formal code description (description of the input data, contents of the COMMON blocks, functions of the IVA2/001 routines). In addition the nonformal description of the current IVA2/001 constitutive package and the reactor core model are given. (orig.) [de

  10. Fuel assembly for pressure loss variable PWR type reactor

    International Nuclear Information System (INIS)

    Yoshikuni, Masaaki.

    1993-01-01

    In a PWR type reactor, a pressure loss control plate is attached detachably to a securing screw holes on the lower surface of a lower nozzle to reduce a water channel cross section and increase a pressure loss. If a fuel assembly attached with the pressure loss control plate is disposed at a periphery of the reactor core where the power is low and heat removal causes no significant problem, a flowrate at the periphery of the reactor core is reduced. Since this flowrate is utilized for removal of heat from fuel assemblies of high powder at the center of the reactor core where a pressure loss control plate is not attached, a thermal limit margin of the whole reactor core is increased. Thus, a limit of power peaking can be moderated, to obtain a fuel loading pattern improved with neutron economy. (N.H.)

  11. Advanced accumulator for PWR

    International Nuclear Information System (INIS)

    Ichimura, Taiki; Chikahata, Hideyuki

    1997-01-01

    Advanced accumulators have been incorporated into the APWR design in order to simplify the safety system configuration and to improve reliability. The advanced accumulators refill the reactor vessel with a large discharge flow rate in a large LOCA, then switch to a small flow rate to continue safety injection for core reflooding. The functions of the conventional accumulator and the low head safety injection pump are integrated into this advanced accumulator. Injection performance tests simulating LOCA conditions and visualization tests for new designs have been carried out. This paper describes the APWR ECCS configuration, the advanced accumulator design and some of the injection performance and visualization test results. It was verified that the flow resistance of the advanced accumulator is independent of the model scale. The similarity law and performance data of the advanced accumulator for applying APWR was established. (author)

  12. 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

  13. Embedded EZ-Source Inverters

    DEFF Research Database (Denmark)

    Blaabjerg, Frede; Loh, Poh Chiang; Gao, F.

    2008-01-01

    -voltage oscillations to the system. Therefore, Z-source inverters are in effect safer and less complex, and can be implemented using only passive elements with no additional active semiconductor needed. Believing in the prospects of Z-source inverters, this paper contributes by introducing a new family of embedded EZ...

  14. Embedded EZ-Source Inverters

    DEFF Research Database (Denmark)

    Chiang Loh, Poh; Gao, Feng; Blaabjerg, Frede

    2010-01-01

    overvoltage oscillations to the system. Therefore, Z-source inverters are, in effect, safer and less complex and can be implemented using only passive elements with no additional active semiconductor needed. Believing in the prospects of Z-source inverters, this paper contributes by introducing a new family...

  15. Simplified High-Power Inverter

    Science.gov (United States)

    Edwards, D. B.; Rippel, W. E.

    1984-01-01

    Solid-state inverter simplified by use of single gate-turnoff device (GTO) to commutate multiple silicon controlled rectifiers (SCR's). By eliminating conventional commutation circuitry, GTO reduces cost, size and weight. GTO commutation applicable to inverters of greater than 1-kilowatt capacity. Applications include emergency power, load leveling, drives for traction and stationary polyphase motors, and photovoltaic-power conditioning.

  16. 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)

  17. 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

  18. Electronic inverter assembly

    Science.gov (United States)

    Singh, Brij N.; Schmit, Christopher J.

    2018-05-22

    A first driver portion comprises a set of first components mounted on or associated with a first circuit board. A second circuit board is spaced apart from the first circuit board. A second driver portion comprises a set of second components mounted on or associated with the second circuit board, where the first driver portion and the second driver portion collectively are adapted to provide input signals to the control terminal of each semiconductor switch of an inverter. A first edge connector is mounted on the first circuit board. A second edge connector is mounted on the second circuit board. An interface board has mating edges that mate with the first edge connector and the second edge connector.

  19. The one-dimensional normalised generalised equivalence theory (NGET) for generating equivalent diffusion theory group constants for PWR reflector regions

    International Nuclear Information System (INIS)

    Mueller, E.Z.

    1991-01-01

    An equivalent diffusion theory PWR reflector model is presented, which has as its basis Smith's generalisation of Koebke's Equivalent Theory. This method is an adaptation, in one-dimensional slab geometry, of the Generalised Equivalence Theory (GET). Since the method involves the renormalisation of the GET discontinuity factors at nodal interfaces, it is called the Normalised Generalised Equivalence Theory (NGET) method. The advantages of the NGET method for modelling the ex-core nodes of a PWR are summarized. 23 refs

  20. 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)

  1. PWR fuel management optimization

    International Nuclear Information System (INIS)

    Dumas, Michel.

    1981-10-01

    This report is aimed to the optimization of the refueling pattern of a nuclear reactor. At the beginning of a reactor cycle a batch of fuel assemblies is available: the physical properties of the assemblies are known: the mathematical problem is to determine the refueling pattern which maximizes the reactivity or which provides the flattest possible power distribution. The state of the core is mathematically characterized by a system of partial derivative equations, its smallest eigenvalue and the associated eigenvector. After a study of the convexity properties of the problem, two algorithms are proposed. The first one exhanges assemblies to improve the starting configurations. The enumeration of the exchanges is limited to the 2 by 2, 3 by 3, 4 by 4 permutations. The second one builds a solution in two steps: in the first step the discrete variables are replaced by continuous variables. The non linear optimization problem obtained is solved by ''the Method of Approximation Programming'' and in the second step, the refuelling pattern which provides the best approximation of the optimal power distribution is searched by a Branch an d Bound Method [fr

  2. Analisis Harmonisa Inverter PWM Satu Fasa

    OpenAIRE

    Rejeki Simanjorang

    2008-01-01

    Pada tesis ini dianalisis harmonisa inverter PWM satu fasa. Inverter PWM satu fasa yang akan ditinjau adalah inverter satu fasa jembatan penuh (konvensional) dan inverter komposit. Analisis difokuskan pada penentuan pola penyaklaran yang optimum agar pembangkitan harmonisa dan switching losses inverter rendah. Untuk menentukan pola penyaklaran optimum maka dilakukan analisis yang berbasis pada rangkaian ekivalen harmonisa inverter satu fasa. Dengan menggunakan pola penyaklaran optimum, kedua ...

  3. Transformerless PV inverters. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Borup, U.

    2009-12-15

    Since the start of the project the market for grid connected PV inverters have developed further. When the project started three - phase inverter were only available in high power systems. The technology developed within this project will enable three phase technology also to be implemented in string inverters for system down to 10 kW. We expect this to be very attractive due to the increased demand for symmetrical feed-in to the grid. The project relevance is therefore high and the sector continues to develop very much driven by technology. Especially the inverter technology is getting a lot of focus. The inverter systems are expected to take a much larger role in supporting the electrical grid in the future. The technology platform developed within the project is prepared to be extended with these utility functionalities. The main results of the project were: 1) A new technology concept for transformer-less inverters has been demonstrated with a number of prototypes. 2) Efficiency above 97,7% has been proven. 3) Efficiency and Maximum power point tracking has been optimized to ensure that almost all energy produced of the panels is transferred to the grid. 4) The platform is developed with a very fast control board, which enables extended functionality as demanding grid supporting functions in the future. Details about cost price and details about the control loop implementation is excluded from the report due to the competitive situation for Danfoss Solar Inverters A/S. (LN)

  4. Application of the BEACON-TSM system to the operation of PWR reactors

    International Nuclear Information System (INIS)

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

    2011-01-01

    BEACON-TSM is an advanced system of the operation support of PWR reactors that combines the capabilities of an advanced nodal neutronic model and the measures of the instrumentation available in plant to determine, accurately and continuously, the distribution of power in the core and the available margins to the limits of the beak factors.

  5. Fuel Cycle Cost Calculations for a 120,000 shp PWR for Ship Propulsion. RCN Report

    International Nuclear Information System (INIS)

    Dekker, N.H.; Foggi, C.; Giacomazzi, G.

    1972-02-01

    A parametric study of the fuel cycle costs for a 120,000 SHP PWR for ship propulsion has been carried out. Variable parameters are: fuel pellet diameter, moderating ratio and refuelling scheme. Minimum fuel cycle costs can be obtained at moderating ratios of about 2.2. Both fuel cycle costs and reactor control requirements favour the two batch core. (author)

  6. A simulated test of physical starting and reactor physics on zero power facility of PWR

    International Nuclear Information System (INIS)

    Yao Zewu; Ji Huaxiang; Chen Zhicheng; Yao Zhiquan; Chen Chen; Li Yuwen

    1995-01-01

    The core neutron economics has been verified through experiments conducted at a zero power reactor with baffles of various thickness. A simulated test of physical starting of Qinshan PWR has been introduced. The feasibility and safety of the programme are verified. The research provides a valuable foundation for developing physical starting programme

  7. Modulation of Current Source Inverter

    Directory of Open Access Journals (Sweden)

    Golam Reza Arab Markadeh

    2011-04-01

    Full Text Available Direct torque control with Current Source Inverter (CSI instead of voltage source inverter is so appropriate because of determining the torque of induction motor with machine current and air gap flux. In addition, Space-Vector Modulation (SVM is a more proper method for CSI because of low order harmonics reduction, lower switching frequency and easier implementation. This paper introduces the SVM method for CSI and uses the proposed inverter for vector control of an induction motor. The simulation results illustrate fast dynamic response and desirable torque and speed output. Fast and accurate response to changes of speed and load torque reference completely proves the prominence of this method.

  8. Detecting and mitigating inverter aging

    International Nuclear Information System (INIS)

    Gunther, W.E.; Taylor, J.H.; Aggarwal, S.K.

    1988-01-01

    Nuclear power plants use inverters to supply power to safety-related equipment, instrumentation, and controls. They convert direct current (dc) to alternating current (ac) power, thereby making low voltage ac power available even under a station blackout condition. As part of the U.S. NRC's nuclear plant aging research (NPAR) program, the operating experience of this equipment has been analyzed to determine the dominant failure modes and causes. This paper summarizes that data, and then describes methods which can be employed to detect inverter degradation prior to failure, as well as methods to minimize the failure effects. In both cases, the mitigation of inverter aging is emphasized

  9. Plutonium recycle in PWR reactors (Brazilian Nuclear Program)

    International Nuclear Information System (INIS)

    Rubini, L.A.

    1978-02-01

    An evaluation is made of the material requirements of the nuclear fuel cycle with plutonium recycle. It starts from the calculation of a reference reactor and allows the evaluation of demand under two alternatives of nuclear fuel cycle for Pressurized Water Reactors (PWR): without plutonium recycle; and with plutonium recycle. Calculations of the reference reactor have been carried out with the CELL-CORE codes. For plutonium recycle, the concept of uranium and plutonium homogeneous mixture has been adopted, using self-produced plutonium at equilibrium, in order to get minimum neutronic perturbations in the reactor core. The refueling model studied in the reference reactor was the 'out-in' scheme with a constant number of changed fuel elements (approximately 1/3 of the core). Variations in the material requirements were studied considering changes in the installed nuclear capacity of PWR reactors, the capacity factor of these reactors, and the introduction of fast breeders. Recycling plutonium produced inside the system can reach economies of about 5%U 3 O 8 and 6% separative work units if recycle is assumed only after the 5th operation cycle of the thermal reactors. The cumulative amount of fissile plutonium obtained by the Brazilian Nuclear Program of PWR reactors by 1991 should be sufficient for a fast breeder with the same capacity as Angra 2. For the proposed fast breeder programs, the fissile plutonium produced by thermal reactors is sufficient to supply fast breeder initial necessities. Howewer, U 3 O 8 and SWU economy with recycle is not significant when the proposed fast breeder program is considered. (Author) [pt

  10. Model for calculating the boron concentration in PWR type reactors

    International Nuclear Information System (INIS)

    Reis Martins Junior, L.L. dos; Vanni, E.A.

    1986-01-01

    A PWR boron concentration model has been developed for use with RETRAN code. The concentration model calculates the boron mass balance in the primary circuit as the injected boron mixes and is transported through the same circuit. RETRAN control blocks are used to calculate the boron concentration in fluid volumes during steady-state and transient conditions. The boron reactivity worth is obtained from the core concentration and used in RETRAN point kinetics model. A FSAR type analysis of a Steam Line Break Accident in Angra I plant was selected to test the model and the results obtained indicate a sucessfull performance. (Author) [pt

  11. 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)

  12. 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)

  13. Estimating probable flaw distributions in PWR steam generator tubes

    International Nuclear Information System (INIS)

    Gorman, J.A.; Turner, A.P.L.

    1997-01-01

    This paper describes methods for estimating the number and size distributions of flaws of various types in PWR steam generator tubes. These estimates are needed when calculating the probable primary to secondary leakage through steam generator tubes under postulated accidents such as severe core accidents and steam line breaks. The paper describes methods for two types of predictions: (1) the numbers of tubes with detectable flaws of various types as a function of time, and (2) the distributions in size of these flaws. Results are provided for hypothetical severely affected, moderately affected and lightly affected units. Discussion is provided regarding uncertainties and assumptions in the data and analyses

  14. Improvement in PWR flexibility the french program 1975-1995

    International Nuclear Information System (INIS)

    Gautier, A.; Miossec, C.

    1985-12-01

    Between 1975 and 1985, a substantial effort was launched in France to greatly improve PWR's flexibility, resulting in the current situation where both frequency control and load follow are now routinely performed on most plants in operation. Based on rapidly accumulating operational experience and on all expertise acquired in the past decade, a second-generation core control strategy is now being finalized for application on all future 1400 MW plants (with commercial operation scheduled in 1992 for first unit). This 20-year program is discussed

  15. Full MOX core for PWRs

    International Nuclear Information System (INIS)

    Puill, A.; Aniel-Buchheit, S.

    1997-01-01

    Plutonium management is a major problem of the back end of the fuel cycle. Fabrication costs must be reduced and plant operation simplified. The design of a full MOX PWR core would enable the number of reactors devoted to plutonium recycling to be reduced and fuel zoning to be eliminated. This paper is a contribution to the feasibility studies for achieving such a core without fundamental modification of the current design. In view of the differences observed between uranium and plutonium characteristics it seems necessary to reconsider the safety of a MOX-fuelled PWR. Reduction of the control worth and modification of the moderator density coefficient are the main consequences of using MOX fuel in a PWR. The core reactivity change during a draining or a cooling is thus of prime interest. The study of core global draining leads to the following conclusion: only plutonium fuels of very poor quality (i.e. with low fissile content) cannot be used in a 900 MWe PWR because of a positive global voiding reactivity effect. During a cooling accident, like an spurious opening of a secondary-side valve, the hypothetical return to criticality of a 100% MOX core controlled by means of 57 control rod clusters (made of hafnium-clad B 4 C rods with a 90% 10 B content) depends on the isotopic plutonium composition. But safety criteria can be complied with for all isotopic compositions provided the 10 B content of the soluble boron is increased to a value of 40%. Core global draining and cooling accidents do not present any major obstacle to the feasibility of a 100% MOX PWR, only minor hardware modifications will be required. (author)

  16. 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.)

  17. Peaking-factor of PWR

    International Nuclear Information System (INIS)

    Morioka, Noboru; Kato, Yasuji; Yokoi, M.

    1975-01-01

    Output peaking factor often plays an important role in the safety and operation of nuclear reactors. The meaning of the peaking factor of PWRs is categorized into two features or the peaking factor in core (FQ-core) and the peaking factor on the basis of accident analysis (or FQ-limit). FQ-core is the actual peaking factor realized in nuclear core at the time of normal operation, and FQ-limit should be evaluated from loss of coolant accident and other abnormal conditions. If FQ-core is lower than FQ-limit, the reactor may be operated at full load, but if FQ-core is larger than FQ-limit, reactor output should be controlled lower than FQ-limit. FQ-core has two kinds of values, or the one on the basis of nuclear design, and the other actually measured in reactor operation. The first FQ-core should be named as FQ-core-design and the latter as FQ-core-measured. The numerical evaluation of FQ-core-design is as follows; FQ-core-design of three-dimensions is synthesized with FQ-core horizontal value (X-Y) and FQ-core vertical value, the former one is calculated with ASSY-CORE code, and the latter one with one dimensional diffusion code. For the evaluation of FQ-core-measured, on-site data observation from nuclear reactor instrumentation or off-site data observation is used. (Iwase, T.)

  18. Nanowire NMOS Logic Inverter Characterization.

    Science.gov (United States)

    Hashim, Yasir

    2016-06-01

    This study is the first to demonstrate characteristics optimization of nanowire N-Channel Metal Oxide Semiconductor (NW-MOS) logic inverter. Noise margins and inflection voltage of transfer characteristics are used as limiting factors in this optimization. A computer-based model used to produce static characteristics of NW-NMOS logic inverter. In this research two circuit configuration of NW-NMOS inverter was studied, in first NW-NMOS circuit, the noise margin for (low input-high output) condition was very low. For second NMOS circuit gives excellent noise margins, and results indicate that optimization depends on applied voltage to the inverter. Increasing gate to source voltage with (2/1) nanowires ratio results better noise margins. Increasing of applied DC load transistor voltage tends to increasing in decreasing noise margins; decreasing this voltage will improve noise margins significantly.

  19. 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

  20. AREVA solutions to licensing challenges in PWR and BWR reload and safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Curca-Tivig, Florin [AREVA GmbH, Erlangen (Germany)

    2016-05-15

    Regulatory requirements for reload and safety analyses are evolving: new safety criteria, request for enlarged qualification databases, statistical applications, uncertainty propagation.. In order to address these challenges and access more predictable licensing processes, AVERA is implementing consistent code and methodology suites for PWR and BWR core design and safety analysis, based on first principles modeling and extremely broad verification and validation data base. Thanks to the high computational power increase in the last decades methods' development and application now include new capabilities. An overview of the main AREVA codes and methods developments is given covering PWR and BWR applications in different licensing environments.

  1. Design of a steam generator for PWR power plants and steady state simulation

    International Nuclear Information System (INIS)

    Ferreira, W.J.

    1982-01-01

    A procedure and a computer code for the thermal design of a steam generator for PWR power plants is developed. A vertical integral steam generator with inverted U-tubes and natural circulation of the secondary side is selected for modelling. Primary fluid velocity and recirculation ratio are varied to obtain the preliminary dimensions. Further, adjustments are made through iteractive solution of the equations of conservation of mass, energy and momentum. An agreement is found between design calculations for steam generators of different capacities and existing designs. (Author) [pt

  2. In-core fuel management activities in China

    International Nuclear Information System (INIS)

    Ruan Keqiang; Chen Renji; Hu Chuanwen

    1990-01-01

    The development of nuclear power in China has reached such a stage that PWR in-core fuel management becomes an urgent problem. At present the main effort is concentrated on solving the Qinshan nuclear power plant and Daya Bay nuclear power plant fuel management problems. For the Qinshan PWR (300 MWe) two packages of in-core fuel management code were developed, one with simplified nodal diffusion method and the other uses advanced Green's function nodal method. Both were used in the PWR core design. With the help of the two code packages first two cycles of the Qinshan PWR core burn-up were calculated. Besides, several research works are under way in the following areas: improvement of the nodal diffusion method and other coarse mesh method in terms of computing speed and accuracy; backward diffusion technique for fuel management application; optimization technique in the fuel loading pattern searching. As for the Daya Bay PWR plant (twin 900 MWe unit), the problem about using what kind of code package for in-core fuel management is still under discussion. In principle the above mentioned code packages are also applicable to it. Besides PWR, in-core fuel management research works are also under way for research reactors, for example, heavy water research reactor and high flux research reactor in some institutes in China. China also takes active participation in international in-core fuel management activities. (author). 19 refs

  3. Voltage-spike analysis for a free-running parallel inverter

    Science.gov (United States)

    Lee, F. C. Y.; Wilson, T. G.

    1974-01-01

    Unwanted and sometimes damaging high-amplitude voltage spikes occur during each half cycle in many transistor saturable-core inverters at the moment when the core saturates and the transistors switch. The analysis shows that spikes are an intrinsic characteristic of certain types of inverters even with negligible leakage inductance and purely resistive load. The small but unavoidable after-saturation inductance of the saturable-core transformer plays an essential role in creating these undesired thigh-voltage spikes. State-plane analysis provides insight into the complex interaction between core and transistors, and shows the circuit parameters upon which the magnitude of these spikes depends.

  4. Method of core thermodynamic reliability determination in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ackermann, G.; Horche, W. (Ingenieurhochschule Zittau (German Democratic Republic). Sektion Kraftwerksanlagenbau und Energieumwandlung)

    1983-01-01

    A statistical model appropriate to determine the thermodynamic reliability and the power-limiting parameter of PWR cores is described for cases of accidental transients. The model is compared with the hot channel model hitherto applied.

  5. Method of core thermodynamic reliability determination in pressurized water reactors

    International Nuclear Information System (INIS)

    Ackermann, G.; Horche, W.

    1983-01-01

    A statistical model appropriate to determine the thermodynamic reliability and the power-limiting parameter of PWR cores is described for cases of accidental transients. The model is compared with the hot channel model hitherto applied. (author)

  6. Analyses of plant behaviors at the secondary side depressurization during LOCA of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kawabe, Yasuharu; Tamaki, Tomohiko; Kohriyama, Tamio; Ohtani, Masanori [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2001-09-01

    When high pressure injection systems failed during a small break loss-of-coolant-accident (LOCA) for a PWR, main steam relief valves are opened to operate accumulator systems. However, it is pointed out that the core can be exposed since so-called counter current flow limitation (CCFL) occurs in steam generator (SG) tubes. The possibility of the core exposure by CCFL in a PWR plant was evaluated. First, RELAP5/MOD2 code was modified to be able to calculate CCFL. And then the code was applied to evaluate a 4-loop PWR plant. The LOCA with a rupture 3 inches were analyzed with the following two cases: (1) Only the main steam relief valve of the loop with the rupture is opened. (2) all of the relief valves are opened. It is seen that the CCFL phenomenon occurs in the case (1), however, the core cooling was maintained by the accumulator systems that actuated during the core exposure. On the other hand, the core exposure by CCFL is not observed in the case (2). It is shown that core cooling is promoted by operation of main steam relief valves. (author)

  7. VALIDATION OF SIMBAT-PWR USING STANDARD CODE OF COBRA-EN ON REACTOR TRANSIENT CONDITION

    Directory of Open Access Journals (Sweden)

    Muhammad Darwis Isnaini

    2016-03-01

    Full Text Available The validation of Pressurized Water Reactor typed Nuclear Power Plant simulator developed by BATAN (SIMBAT-PWR using standard code of COBRA-EN on reactor transient condition has been done. The development of SIMBAT-PWR has accomplished several neutronics and thermal-hydraulic calculation modules. Therefore, the validation of the simulator is needed, especially in transient reactor operation condition. The research purpose is for characterizing the thermal-hydraulic parameters of PWR1000 core, which be able to be applied or as a comparison in developing the SIMBAT-PWR. The validation involves the calculation of the thermal-hydraulic parameters using COBRA-EN code. Furthermore, the calculation schemes are based on COBRA-EN with fixed material properties and dynamic properties that calculated by MATPRO subroutine (COBRA-EN+MATPRO for reactor condition of startup, power rise and power fluctuation from nominal to over power. The comparison of the temperature distribution at nominal 100% power shows that the fuel centerline temperature calculated by SIMBAT-PWR has 8.76% higher result than COBRA-EN result and 7.70% lower result than COBRA-EN+MATPRO. In general, SIMBAT-PWR calculation results on fuel temperature distribution are mostly between COBRA-EN and COBRA-EN+MATPRO results. The deviations of the fuel centerline, fuel surface, inner and outer cladding as well as coolant bulk temperature in the SIMBAT-PWR and the COBRA-EN calculation, are due to the value difference of the gap heat transfer coefficient and the cladding thermal conductivity.

  8. Γ-source Neutral Point Clamped Inverter

    DEFF Research Database (Denmark)

    Mo, Wei; Loh, Poh Chiang; Blaabjerg, Frede

    Transformer based Z-source inverters are recently proposed to achieve promising buck-boost capability. They have improved higher buck-boost capability, smaller size and less components count over Z-source inverters. On the other hand, neutral point clamped inverters have less switching stress...... and better output performance comparing with traditional two-level inverters. Integrating these two types of configurations can help neutral point inverters achieve enhanced votlage buck-boost capability....

  9. A study of 2-Dimensional effects in the core of a PWR during the refloading phase of a LOCA. Analysis of data of PERICLES experiments with the COBRA-NC code

    International Nuclear Information System (INIS)

    Reinhardt, H.J.

    1989-09-01

    The project is embedded in the Shared Cost Action Programme (SCA) of the European Communities (CEC) on Reactor Safety, Research Area No. 4, concerning the analysis of experimental data on loss-of-coolant accidents and emergency core cooling. The PERICLES experiments, performed at CEA in Grenoble, had the objective to study multidimensional effects under well defined conditions concentrating on the inter-assembly character of reflood phenomena. The general aim of the present project is to analyse PERICLES experimental data in order to improve models in relevant system codes. Particular objectives of the project are - the critical evaluation of the experimental data of PERICLES Run 8; - the drawing of conclusions from the data with respect to physical and geometrical models for the multi-bundle reflood analysis; - the performance of one-and multi-dimensional computations with COBRA-NC; - the comparison of computational and experimental data; and - the development of conclusions and specifications of additional research needed. The analysis of the experimetal data of Run 8 was performed by a computer programme developed for postprocessing data of any PERICLES experiment. The postprocessor includes an automatic location of the quenchfront and displays it graphically with respect to time, vertical and horizontal directions. Furthermore, rod and fluid temperatures versus height, quenchtimes versus height, densities versus height, and temperatures, pressures, densities etc. versus time can be plotted. As far as computer simulations are concerned, it was one of the objectives of the present study to analyse in greater detail the multidimensional phenomena during the reflooding phase of a LOCA and to compare the numerical results with the experimental data. Such simulation may serve to adjust and improve existing computer codes as well as to validate the codes. Moreover, computer simulations are able to give information which are not available from experimental data; in the

  10. A comparison of in-vessel behaviors between SFR and PWR under severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sanggil; Cho, Cheon Hwey [ACT Co., Daejeon (Korea, Republic of); Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    This paper aims to provide an easy guide for experts who know well the severe accident phenomenology of Pressurized Water Reactor (PWR) by comparing both reactor design concepts and in vessel behaviors under a postulated severe accident condition. This study only provides a preliminary qualitative comparison based on available literature. The PWR and SFR in-vessel design concepts and their effects under a postulate severe accident are investigated in this paper. Although this work is a preliminary study to compare the in-vessel behaviors for both PWR and SFR, it seems that there is no possibility to lead a significant core damage in the metal fuel SFR concept. In the oxide fuel SFR, there might be a chance to progress to the severe accident initiators such as the energetic reaction, flow blockage and so on.

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

    International Nuclear Information System (INIS)

    Greneche, D.

    2014-01-01

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

  12. Study on advanced nuclear fuel cycle of PWR/CANDU synergism

    International Nuclear Information System (INIS)

    Xie Zhongsheng; Huo Xiaodong

    2002-01-01

    According to the concrete condition that China has both PWR and CANDU reactors, one of the advanced nuclear fuel cycle strategy of PWR/CANDU synergism ws proposed, i.e. the reprocessed uranium of spent PWR fuel was used in CANDU reactor, which will save the uranium resource, increase the energy output, decrease the quantity of spent fuels to be disposed and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, the transition from the natural uranium to the recycled uranium (RU) can be completed without any changes of the structure of reactor core and operation mode. Furthermore, because of the low radiation level of RU, which is acceptable for CANDU reactor fuel fabrication, the present product line of fuel elements of CANDU reactor only need to be shielded slightly, also the conditions of transportation, operation and fuel management need not to be changed. Thus this strategy has significant practical and economical benefit

  13. Inverted emulsion drilling fluid

    Energy Technology Data Exchange (ETDEWEB)

    Ana, I; Astanei, E; Mireanu, G; Orosz, M; Popescu, F; Vasile, I

    1979-07-28

    The subject of the invention is the method of obtaining inverted drilling fluid which is required during stripping of a productive bed and ending of a well where difficulties develop during drilling of the argillaceous rock. Example: in a reservoir with capacity 30 m/sup 3/, 10 m/sup 3/ of diesel fuel are added. A total of 1000 kg of emulsifier are added to the diesel fuel consisting of: 85 mass% of a mixture of sodium and potassium salts of fatty acids, residues of fatty acids or naphthene acids with high molecular weight taken in proportion of 10:90; 5 mass% of a mixture of polymers with hydrophilic-hydrophobic properties obtained by mixing 75 mass% of polyethylene oxide with molecular weight 10,000 and 25 mass% of propylene oxide with molecular weight 15,000, and 10 mass% of salt on alkaline earth metal (preferably calcium chloride). The mixture is mixed into complete dissolving. Then 1200 kg of filtering accelerator are added obtained from concentrated sulfuric acid serving for sulfur oxidation, asphalt substance with softening temperature 85-104/sup 0/C and fatty acids C/sub 10/-C/sub 20/ taken in a proportion of 23.70 and 7 mass% The mixture obtained in this manner is neutralized by adding calcium hydroxide and equal quantities of alumina and activated bentonite clay in a concentration of 1-10 mass%, more preferably 5 mass% in relation to the initial mixture. The obtained mass is mixed until complete dispersion, after which 200 kg of organophilic clay are added obtained from bentonite of the type montmorillonite of sodium by processing with derivate obtained from amine of the type of the quaternary base of ammonium salt, and agent of hydrophobization of the type of fatty alcohols, fatty acids, nonion surfactants of the block-polymer type. After complete dispersion of the organophilic clay, 100 kg of stabilizer of emulsion of the surfactant type was added with molecular weight of 250010,000, more preferably 5000, in concentration of 0.1-5.0 mass%, more

  14. PWR type process heat reactor

    International Nuclear Information System (INIS)

    Aubert, Gilles; Petit, Guy.

    1974-01-01

    The nuclear reactor described is of the pressurized water type. It includes a prestressed concrete vessel, the upper part of which is shut by a closure, and a core surrounded by a core ring. The core fuel assemblies are supported by an initial set of vertical tubes integral with the bottom of the vessel, which serve to guide the rods of the control system. Over the core there is a second set of vertical tubes, able to receive the absorbing part of a control rod when this is raised above the core. An annular pressurizer around the core ring keeps the water in a liquid state. A pump is located above the second set of tubes and is integral with the closure. It circulates the water between the core and the intake of at least one primary heat exchanger, the exchanger (s) being placed between the wall of the vessel and the core ring [fr

  15. Preliminary safety analysis of the PWR with accident-tolerant fuels during severe accident conditions

    International Nuclear Information System (INIS)

    Wu, Xiaoli; Li, Wei; Wang, Yang; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng; Liu, Tong; Deng, Yongjun; Huang, Heng

    2015-01-01

    Highlights: • Analysis of severe accident scenarios for a PWR fueled with ATF system is performed. • A large-break LOCA without ECCS is analyzed for the PWR fueled with ATF system. • Extended SBO cases are discussed for the PWR fueled with ATF system. • The accident-tolerance of ATF system for application in PWR is illustrated. - Abstract: Experience gained in decades of nuclear safety research and previous nuclear accidents direct to the investigation of passive safety system design and accident-tolerant fuel (ATF) system which is now becoming a hot research point in the nuclear energy field. The ATF system is aimed at upgrading safety characteristics of the nuclear fuel and cladding in a reactor core where active cooling has been lost, and is preferable or comparable to the current UO 2 –Zr system when the reactor is in normal operation. By virtue of advanced materials with improved properties, the ATF system will obviously slow down the progression of accidents, allowing wider margin of time for the mitigation measures to work. Specifically, the simulation and analysis of a large break loss of coolant accident (LBLOCA) without ECCS and extended station blackout (SBO) severe accident are performed for a pressurized water reactor (PWR) loaded with ATF candidates, to reflect the accident-tolerance of ATF

  16. PSA LEVEL 3 DAN IMPLEMENTASINYA PADA KAJIAN KESELAMATAN PWR

    Directory of Open Access Journals (Sweden)

    Pande Made Udiyani

    2015-03-01

    series of calculations performed are: calculate the source terms of the core damaged, modeling of meteorological conditions and environmental site, exposure pathway modeling, analysis of radionuclide dispersion and transport phenomena in the environment, radionuclide deposition analysis, analysis of radiation dose, protection & mitigation analysis, and risk analysis. The assessment uses a series of subsystems on PC Cosyma software. The results prove that the safety assessment using Level 3 PSA methodology is very effective and comprehensive estimate the impact, consenquences, risks, nuclear emergency preparedness, and the reactor accident management especially for severe accidents or beyond design basis accidents of nuclear power plants. The results of the assessment can be used as a feedback to safety assessment of Level 1 PSA and Level 2 PSA. Keywords: Level 3 PSA, accident, PWR

  17. An analysis of transients in the PWR downcomer

    International Nuclear Information System (INIS)

    Jovanovic, A.

    1981-01-01

    The paper deals with the problem of determining non-stationary temperature field in the downcomer of a PWR type reactor. For this purpose, an analytical model has been developed. The model covers five components of (PWR - Krsko) downcomer: the core-barrel, floor between the core-barrel and the thermal shield, the thermal shield, flow between the thermal shield and the reactor vessel wall, the reactor vessel wall. The model includes internal heat generation in metal structures. The governing equations of the model have been written in the finite difference explicit form. The system of resulting algebraic equations was solved bu Gauss-Seidel method, using a modular computer code. Several characteristic transients were examined (step and continuous change of fluid temperature at the inlet nozzle). Also, an analysis of main parameters (heat transfer coefficient and flow rate) has been performed. The model is intended to be used as basics for further development of a more realistic model that could be used for practical safety analysis. (author)

  18. Preparation of inverted medium and processing in the inverted medium

    International Nuclear Information System (INIS)

    Zafarullah, Ijaz; Tian Minghzen; Chang Tiejun; Randall Babbitt, W.

    2007-01-01

    The processing of weak optical signals in spatial-spectral holographic (SSH) materials coherently inverted with optical frequency chirped pulses were investigated. Simulations and experimental studies in Tm 3+ :YAG were conducted to characterize the parameters of the frequency chirped laser pulse used to invert the SSH material in order to obtain high photon echo efficiency for SSH lidar processing. Collinear and angled beam geometries and single shot and accumulated processes were investigated. Echo efficiencies as high as 450% were measured, significantly higher than the typical stimulated photon echo efficiency of 10%

  19. Preparation of inverted medium and processing in the inverted medium

    Energy Technology Data Exchange (ETDEWEB)

    Zafarullah, Ijaz [Department of Physics, Montana State University, Bozeman, MT 59717 (United States)]. E-mail: ijaz@montana.edu; Tian Minghzen [Department of Physics, Montana State University, Bozeman, MT 59717 (United States); Chang Tiejun [The Spectrum Laboratory, Montana State University, Bozeman, MT 59717 (United States); Randall Babbitt, W. [Department of Physics, Montana State University, Bozeman, MT 59717 (United States)

    2007-11-15

    The processing of weak optical signals in spatial-spectral holographic (SSH) materials coherently inverted with optical frequency chirped pulses were investigated. Simulations and experimental studies in Tm{sup 3+}:YAG were conducted to characterize the parameters of the frequency chirped laser pulse used to invert the SSH material in order to obtain high photon echo efficiency for SSH lidar processing. Collinear and angled beam geometries and single shot and accumulated processes were investigated. Echo efficiencies as high as 450% were measured, significantly higher than the typical stimulated photon echo efficiency of 10%.

  20. Design study of a PWR of 1.300 MWe of Angra-2 type operating in the thorium cycle

    International Nuclear Information System (INIS)

    Andrade, E.P.; Carneiro, F.A.N.; Schlosser, G.J.

    1984-01-01

    The utilization of the thorium-highly enriched uranium and thorium-plutonium mixed oxide fuels in an unmodified PWR is analysed. The PWR of 1300 MWe from KWU (Angra-2 type) is taken as the reference reactor for the study. Reactor core design calculations for both types of fuels considering once-through and recycle fuels. The calculations were performed with the KWU design codes FASER-3 and MEDIUM 2.2 after introduction of the thorium chain and some addition of nuclide data in FASER-3. A two-energy group scheme and a two-dimensional (XY) representation of the reactor core were utilized. (Author) [pt

  1. Study for identification of control rod drops in PWR reactors at any burnup step

    International Nuclear Information System (INIS)

    Souza, Thiago J.; Martinez, Aquilino S.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C.

    2013-01-01

    The control rod drop event in PWR reactors induces an unsafe operating condition. Therefore, in a scenario of a control rod drop is important to quickly identify the rod to minimize undesirable effects. The objective of this work is to develop an on-line method for identification of control rod drop in PWR reactors. The method consists on the construction of a tool that is based on the ex-core detector responses. Therefore, it is proposed to recognize patterns in the neutron ex-core detectors responses and thus to identify on-line a control rod drop in the core during the reactor operation. The results of the study, as well as the behavior of the detector responses, demonstrated the feasibility of this method. (author)

  2. Study for identification of control rod drops in PWR reactors at any burnup step

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Thiago J.; Martinez, Aquilino S.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C., E-mail: tsouza@nuclear.ufrj.br, E-mail: aquilino@lmp.ufrj.br, E-mail: canedo@lmp.ufrj.br, E-mail: alessandro@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    The control rod drop event in PWR reactors induces an unsafe operating condition. Therefore, in a scenario of a control rod drop is important to quickly identify the rod to minimize undesirable effects. The objective of this work is to develop an on-line method for identification of control rod drop in PWR reactors. The method consists on the construction of a tool that is based on the ex-core detector responses. Therefore, it is proposed to recognize patterns in the neutron ex-core detectors responses and thus to identify on-line a control rod drop in the core during the reactor operation. The results of the study, as well as the behavior of the detector responses, demonstrated the feasibility of this method. (author)

  3. Research and development of in-core transducers at the CIAE

    International Nuclear Information System (INIS)

    Huang Yucai; Liu Yupu; Jia Guozhen; Liu Lianping

    1996-01-01

    In this paper, R and D of in-core transducers at the CIAE are briefly summarized. With the construction and commissioning of PWR nuclear power plant in China, fuel rod behaviour need to be studied carefully. As conventional transducers cannot meet the requirements of in-core applications, R and D of in-core transducers are developed. Since 1980's, several kinds of in-core transducers have been successfully fabricated and tested under the conditions simulating PWR. At present, in-pile tests of the transducers combining with the studies of individual behaviour of PWR fuel rod are being planned at the CIAE. (author). 11 refs, 12 figs, 4 tabs

  4. 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)

  5. Accelerating development of advanced inverters :

    Energy Technology Data Exchange (ETDEWEB)

    Neely, Jason C.; Gonzalez, Sigifredo; Ropp, Michael; Schutz, Dustin

    2013-11-01

    The high penetration of utility interconnected photovoltaic (PV) systems is causing heightened concern over the effect that variable renewable generation will have on the electrical power system (EPS). These concerns have initiated the need to amend the utility interconnection standard to allow advanced inverter control functionalities that provide: (1) reactive power control for voltage support, (2) real power control for frequency support and (3) better tolerance of grid disturbances. These capabilities are aimed at minimizing the negative impact distributed PV systems may have on EPS voltage and frequency. Unfortunately, these advanced control functions may interfere with island detection schemes, and further development of advanced inverter functions requires a study of the effect of advanced functions on the efficacy of antiislanding schemes employed in industry. This report summarizes the analytical, simulation and experimental work to study interactions between advanced inverter functions and anti-islanding schemes being employed in distributed PV systems.

  6. Simplified model of a PWR primary coolant circuit

    International Nuclear Information System (INIS)

    Souza, A.L. de; Faya, A.J.G.

    1988-01-01

    The computer program RENUR was developed to perform a very simplified simulation of a typical PWR primary circuit. The program has mathematical models for the thermal-hydraulics of the reactor core and the pressurizer, the rest of the circuit being treated as a single volume. Heat conduction in the fuel rod is analysed by a nodal model. Average and hot channels are treated so that the bulk response of the core and DNBR can be evaluated. A Homogenenous model is employed in the pressurizer. Results are presented for a steady-state situation as well as for a loss of load transient. Agreement with the results of more elaborate computer codes is good with substantial reduction in computer costs. (author) [pt

  7. Inverter for Interchangeable Use as Current Source Inverter and Voltage Source Inverter for Interconnecting to Grid

    Science.gov (United States)

    Teruya, Daisuke; Masukawa, Shigeo; Iida, Shoji

    We propose a novel inverter that can be operated either as a Current Source Inverter (CSI) or as a Voltage Source Inverter (VSI) by changing only the control signals. It is proper to apply it to the interconnecting system with renewal energy, such as photovoltaic cells or wind generation systems, to a grid. This inverter is usually operated as the CSI connected to the grid. Even if the energy source has a lower voltage than the grid, the energy can be supplied to the grid through the proposed inverter. The power factor can be briefly maintained at almost unity. When power supply from the grid is interrupted, the proposed circuit should be operated as the VSI in the stand-alone operation mode. In this way, the circuit can maintain a constant output voltage to the loads. In this paper, the proposed circuit configuration and the control schemes for both the CSI and the VSI are described. Further, the circuit characteristics for both are discussed experimentally.

  8. 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

  9. Thermo-mechanical analysis of PWR bolts susceptible to IASCC

    International Nuclear Information System (INIS)

    Matteoli, C.; Hannink, M.H.C.; Blom, F.J.; Marck, S.C. van der; Charpin-Jacobs, F.

    2015-01-01

    Irradiation Assisted Stress Corrosion Cracking (IASCC) is considered a primary ageing issue for the Reactor Pressure Vessel (RPV) internals of Pressurized Water Reactors (PWR). In particular, this complex phenomenon which develops in an environment featuring thermal and mechanical stresses, interaction with corrosive compounds and irradiation, is affecting the bolts connecting the baffles and the formers in the Nuclear Power Plants' RPVs. The baffle-former assembly is the structure that borders the fuel assemblies region, contributing to keep them in position and separating in the radial direction, the core region from the downcomer region. An evaluation of the stresses and temperatures reached in the baffle-former bolts during normal operation was performed by means of a coupled thermo-mechanical study which uses reactor physics calculations to obtain the fluence in the reactor core and as a consequence the heat deposition in the RPV internals. The heat deposition data are coupled with a finite element model of the bolts and the RPV internals in order to perform a complete analysis taking in account thermal, mechanical and radiation loadings. The study is first carried out focusing on a section of the RPV internals, showing a single row of baffle-former bolts. Then the work is extended to the full core height. The model set up in this work, includes an in-depth study of the behavior of the core internals, in particular baffle-former bolts. The model has the capability of understanding the mechanical and thermal behavior of essential internal components in a PWR. (authors)

  10. PWR primary system chemistry control during hot functional testing

    International Nuclear Information System (INIS)

    Reid, Richard D.; Little, Michael J.

    2014-01-01

    Hot Functional Testing (HFT) involves a number of pre-operational exercises performed to confirm the operability of plant systems at conditions expected during both normal and off-normal operation of a pressurized water reactor (PWR), including operability of safety systems. While the primary purposes of HFT are to demonstrate operability of plant systems and satisfy regulatory requirements, chemistry control during HFT is important to long-term integrity and performance of plant systems. Specifically, HFT is the first time plant equipment is exposed to high temperature water and the chemistry maintained during HFT can impact the passivation layers that form on wetted surfaces and long-term release of metals from these surfaces. Metals released from the inner surfaces of steam generator tubing and reactor coolant loop piping become activated in the core and can redeposit on ex-core surfaces. Because HFT is performed before fuel is loaded in the core, HFT provides an opportunity to produce a passive layer on primary surfaces that is free of activated corrosion products, resistant to metals release during subsequent plant operation, and also resistant to incorporation of activated corrosion products (once fuel is loaded in the core). Thus, maintaining desirable primary chemistry control during HFT is important for source term management, minimization of future shutdown activity releases, minimization of dose rates, and asset preservation. This paper presents an overview of passive film formation in the austenitic stainless steel and high nickel alloys that make up the majority of the primary circuit in advanced PWR designs. Based on this information, a summary is provided of the effects on passive film formation of key chemistry parameters that may be controlled during HFT. (author)

  11. Inverter design for high frequency power distribution

    Science.gov (United States)

    King, R. J.

    1985-01-01

    A class of simple resonantly commutated inverters are investigated for use in a high power (100 KW - 1000 KW) high frequency (10 KHz - 20 KHz) AC power distribution system. The Mapham inverter is found to provide a unique combination of large thyristor turn-off angle and good utilization factor, much better than an alternate 'current-fed' inverter. The effects of loading the Mapham inverter entirely with rectifier loads are investigated by simulation and with an experimental 3 KW 20 KHz inverter. This inverter is found to be well suited to a power system with heavy rectifier loading.

  12. An integrated PWR for marine propulsion

    International Nuclear Information System (INIS)

    Letouze, A.; Marecaux, A.; Rollason, J.; Heap, S.; Foster, A.; Jewer, S.; Thompson, A. C.; Williams, A. M.; Beeley, P. A.

    2008-01-01

    Results from a design study for a nuclear propulsion plant utilising a small integrated PWR using many of the inherent safety features of the IRIS design. The design consists of a single pass, low enrichment core housed, together with all associated primary circuit components, within a reactor pressure vessel 10.3 m high and 4.1 m in diameter. Reactor physics calculations were conducted with the codes WIMS9a and MONK8b. The core design contains 21 fuel assemblies each containing 264 UO 2 fuel pins. Each fuel module has a cluster of 24 boron carbide control rods and a central instrumentation channel. The fuel enrichment was 9% in order to achieve the core lifetime requirement of 3000 EFPD at a reactor power of 120 MWth. This gives a discharge burnup of 51,000 MWd/t. To control excess reactivity, two forms of burnable poison are employed: a zirconium dibromide (ZrB 2 ) coating on the fuel compacts, and gadolinium oxide homogeneously mixed in the fuel. Thermal hydraulic calculations were performed using TRAC-P(ND) for steady-state operation and for a number of fault transients. The helical once through steam generators were modelled using heat structure and pipe components and their performance compared to independent calculations including heat transfer correlations for the helical coiled geometry. Intact circuit calculations for steady state were followed by a small break LOCA calculation including the effect of a containment volume which reproduced the gain of coolant effect reported for IRIS. It was demonstrated that the thermal limits were not exceeded for the identified key transients. The dynamic response of the reactor plant to typical power demands was modelled using AcslXtreme software. Several schemes for limiting the power overshoot that was found on rapid increase to full power were examined. It was concluded that the SG must be operated with variable secondary pressure and the best means of reducing power overshoot is to step back the throttle opening

  13. Single stage buck-boost DC-AC neutral point clamped inverter

    DEFF Research Database (Denmark)

    Mo, Wei; Loh, Poh Chiang; Andrew, A.

    2012-01-01

    This paper proposes a new single stage buck-boost DC-AC neutral point clamped inverter topology which integrates the cascaded configurations of recently introduced inductor-capacitor-capacitor-transformer impedance source network (by Adamowicz) and classic NPC configuration. As a consequence......, it has enhanced buck-boost functionality and low output voltage distortions compared to the traditional Z-source inverter; it has continuous input current which reduces the source stress and inverter noise; it also contains two built-in capacitors which can block the DC current in the transformer...... windings thus preventing the core from saturation; lowers the voltage stresses and power losses of inverter switches and reduces the sizes of filtering devices and as well as obtains better output performance compared to the original two-level Z-source inverters. A phase disposition pulse width modulation...

  14. Assessment of TRAC-PF1/MOD1 code for large break LOCA in PWR

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Ohnuki, Akira; Murao, Yoshio; Abe, Yutaka.

    1993-03-01

    As the first step of the REFLA/TRAC code development, the TRAC/PF1/MOD1 code has been assessed for various experiments that simulate postulated large-break loss-of-coolant accident (LBLOCA) in PWR to understand the predictive capability and to identify the problem areas of the code. The assessment calculations were performed for separate effect tests for critical flow, counter current flow, condensation at cold leg and reflood as well as integral tests to understand predictability for individual phenomena. This report summarizes results from the assessment calculations of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The assessment calculations made clear the predictive capability and problem areas of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The areas, listed below, should be improved for more realistic and effective simulation of LBLOCA in PWR: (1) core heat transfer model during blowdown, (2) ECC bypass model at downcomer during refill, (3) condensation model during accumulator injection, and (4) core thermal hydraulic model during reflood. (author) 57 refs

  15. Flow visualization study of inverted annular flow of post dryout heat transfer region

    International Nuclear Information System (INIS)

    Ishii, M.; De Jarlais, G.

    1985-01-01

    The inverted annular flow is important in the area of LWR accident analysis in terms of the maximum cladding temperature and effectiveness of the emergency core cooling. However, the inverted annular flow thermal-hydraulics is not well understood due to its special heat transfer condition of film boiling. In view of this, the inverted flow is studied in detail experimentally. A new experimental apparatus has been constructed in which film boiling heat transfer can be established in a transparent test section. Data on liquid core stability, core break-up mechanism, and dispersed-core liquid slug and droplet sizes are obtained using F 113 as a test fluid. Both high speed movies and flash photographs are used

  16. 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).

  17. Activity incorporation into zinc doped PWR oxides

    International Nuclear Information System (INIS)

    Maekelae, Kari

    1998-01-01

    Activity incorporation into the oxide layers of PWR primary circuit constructional materials has been studied in Halden since 1993. The first zinc injection tests showed that zinc addition resulted in thinner oxide layers on new metal surfaces and reduced further incorporation of activity into already existing oxides. These tests were continued to find out the effects of previous zinc additions on the pickup of activity onto the surface oxides which were subsequently exposed to zinc-free coolant. The results showed that previous zinc addition will continue to reduce the rate of Co-60 build-up on out-of-core surfaces in subsequent exposure to zinc-free coolants. However, the previous Zn free test was performed for a relatively short period of time and the water chemistry programme was continued to find out the long term effects for extended periods without zinc. The activity incorporation into the stainless steel oxides started to increase as soon as zinc dosing to the coolant was stopped. The Co-60 concentration was lowest on all of the coupons which were first oxidised in Zn containing primary coolant. After the zinc injection period the thickness of the oxides increased, but activity in the oxide films did not increase at the same rate. This could indicate that zinc in the oxide blocks the adsorption sites for Co-60 incorporation. The Co-60 incorporation rate into the oxides on Inconel 600 seemed to be linear whether the oxide was pre-oxidised with or without Zn. The results indicate that zinc can either replace or prevent cobalt transport in the oxides. The results show that for zinc injection to be effective it should be carried out continuously. Furthermore the actual mechanism by which Zn inhibits the activity incorporation into the oxides is still not clear. Therefore, additional work has to follow with specified materials to verify the conclusions drawn in this work. (author)

  18. Assessment of erbium as candidate burnable absorber for future PWR operaning cycles: A neutronic and fabrication study

    International Nuclear Information System (INIS)

    Asou, M.; Dehaudt, P.; Porta, J.

    1995-01-01

    Erbium begins to play a role in the control of PWR core reactivity. Generally speaking, burnable absorbers were only used to establish fresh core equilibrium. In France, since the possibility of extending irradiation cycles by 12 to 18 months, then up to 24 and 30 months, has been envisaged, there is renewed interest in burnable absorbers. The fabrication of PWR pellets has been investigated, providing high density and a good erbium homogeneity. The pellets characteristics were consistent with the specifications of PWR fuel. However, with the present process, the grain size remains small. Studies in progress now shows that erbium is not only a valuable alternative to gadolinium, for long fuel cycles (≥18 months) but also a new fuel concept. (orig.)

  19. Inverters for photovoltaics: quality assurance

    International Nuclear Information System (INIS)

    Graf, J.D.; Haeberlin, H.

    2000-01-01

    This project-overview published by the University of Applied Science in Burgdorf, Switzerland, looks back at the history of the university's testing laboratory for inverters for use in photovoltaic installations and discusses the work done there. After its move from Oberburg to Burgdorf in the early nineties, the laboratory was equipped to provide testing facilities for inverters up to 60 kW. Additions have been made to the infrastructure since then to facilitate the testing of various types of inverter produced in Switzerland and other countries. The measurements that can be carried out, including DC-AC conversion efficiency, harmonics, RF interference, islanding, sensitivity to tele-control signals, turn-on power and operating voltage range are described. The improvement in the quality of the inverters tested is discussed and the importance of continued testing as new manufacturers come onto the market is emphasised. The installation of further testing equipment, such as a 25 kW solar generator simulator and apparatus for the testing of maximum-power-tracking efficiency is discussed

  20. Inverting an Introductory Statistics Classroom

    Science.gov (United States)

    Kraut, Gertrud L.

    2015-01-01

    The inverted classroom allows more in-class time for inquiry-based learning and for working through more advanced problem-solving activities than does the traditional lecture class. The skills acquired in this learning environment offer benefits far beyond the statistics classroom. This paper discusses four ways that can make the inverted…

  1. MOX and UOX PWR fuel performances EDF operating experience

    International Nuclear Information System (INIS)

    Provost, Jean-Luc; Debes, Michel

    2005-01-01

    Based on a large program of experimentations implemented during the 90s, the industrial achievement of new FAs designs with increased performances opens up new prospects. The currently UOX fuels used on the 58 EDF PWR units are now authorized up to a maximum FA burn-up of 52 GWd/t with a large experience from 45 to 50 GWd/t. Today, the new products, along with the progress made in the field of calculation methods, still enable to increase further the fuel performances with respect to the safety margins. Thus, the conditions are met to implement in the next years new fuel managements on each NPPs series of the EDF fleet with increased enrichment (up to 4.5%) and irradiation limits (up to 62 GWd/t). The recycling of plutonium is part of EDF's reprocessing/recycling strategy. Up to now, 20 PWR 900 MW reactors are managed in MOX hybrid management. The feedback experience of 18 years of PWR operation with MOX is satisfactory, without any specific problem regarding manoeuvrability or plant availability. EDF is now looking to introduce MOX fuels with a higher plutonium content (up to 8.6%) equivalent to natural uranium enriched to 3.7%. It is the goal of the MOX Parity core management which achieve balance of MOX and UOX fuel performance with a significant increase of the MOX average discharge burn-up (BU max: 52 GWd/t for MOX and UOX). The industrial maturity of new FAs designs, with increased performances, allows the implementation in the next years of new fuel managements on each NPPs series of the EDF fleet. The scheduling of the implementation of the new fuel managements on the PWRs fleet is a great challenge for EDF, with important stakes: the nuclear KWh cost decrease with the improvement of the plant operation performance. (author)

  2. Integral type small PWR with stand-alone safety

    International Nuclear Information System (INIS)

    Makihara, Yoshiaki

    2001-01-01

    A feasibility study is achieved on an integral type small PWR with stand-alone safety. It is designed to have the following features. (1) The coolant does not leak out at any accidental condition. (2) The fuel failure does never occur while it is supposed on the large scale PWR at the design base accident. (3) At any accidental condition the safety is secured without any support from the outside (stand-alone safety secure). (4) It has self-regulating characteristics and easy controllability. The above features can be satisfied by integrate the steam generator and CRDM in the reactor vessel while the pipe line break has to be considered on the conventional PWR. Several counter measures are planned to satisfy the above features. The economy feature is also attained by several simplifications such as (1) elimination of main coolant piping and pressurizer by the integration of primary cooling system and self-pressurizing, (2) elimination of RCP by application of natural circulating system, (3) elimination of ECCS and accumulator by application of static safety system, (4) large scale volume reduction of the container vessel by application of integrated primary cooling system, (5) elimination of boric acid treatment by deletion of chemical shim. The long operation period such as 10 years can be attained by the application of Gd fuel in one batch refueling. The construction period can be shortened by the standardizing the design and the introduction of modular component system. Furthermore the applicability of the reduced modulation core is also considered. (K. Tsuchihashi)

  3. Overload protection system for power inverter

    Science.gov (United States)

    Nagano, S. (Inventor)

    1977-01-01

    An overload protection system for a power inverter utilized a first circuit for monitoring current to the load from the power inverter to detect an overload and a control circuit to shut off the power inverter, when an overload condition was detected. At the same time, a monitoring current inverter was turned on to deliver current to the load at a very low power level. A second circuit monitored current to the load, from the monitoring current inverter, to hold the power inverter off through the control circuit, until the overload condition was cleared so that the control circuit may be deactivated in order for the power inverter to be restored after the monitoring current inverter is turned off completely.

  4. Membrane fusion and inverted phases

    International Nuclear Information System (INIS)

    Ellens, H.; Siegel, D.P.; Alford, D.; Yeagle, P.L.; Boni, L.; Lis, L.J.; Quinn, P.J.; Bentz, J.

    1989-01-01

    We have found a correlation between liposome fusion kinetics and lipid phase behavior for several inverted phase forming lipids. N-Methylated dioleoylphosphatidylethanolamine (DOPE-Me), or mixtures of dioleoylphosphatidylethanolamine (DOPE) and dioleoylphosphatidylcholine (DOPC), will form an inverted hexagonal phase (HII) at high temperatures (above TH), a lamellar phase (L alpha) at low temperatures, and an isotropic/inverted cubic phase at intermediate temperatures, which is defined by the appearance of narrow isotropic 31 P NMR resonances. The phase behavior has been verified by using high-sensitivity DSC, 31 P NMR, freeze-fracture electron microscopy, and X-ray diffraction. The temperature range over which the narrow isotropic resonances occur is defined as delta TI, and the range ends at TH. Extruded liposomes (approximately 0.2 microns in diameter) composed of these lipids show fusion and leakage kinetics which are strongly correlated with the temperatures of these phase transitions. At temperatures below delta TI, where the lipid phase is L alpha, there is little or no fusion, i.e., mixing of aqueous contents, or leakage. However, as the temperature reaches delta TI, there is a rapid increase in both fusion and leakage rates. At temperatures above TH, the liposomes show aggregation-dependent lysis, as the rapid formation of HII phase precursors disrupts the membranes. We show that the correspondence between the fusion and leakage kinetics and the observed phase behavior is easily rationalized in terms of a recent kinetic theory of L alpha/inverted phase transitions. In particular, it is likely that membrane fusion and the L alpha/inverted cubic phase transition proceed via a common set of intermembrane intermediates

  5. The application of neural networks for optimization of the configuration of fuel assemblies in PWR reactors

    International Nuclear Information System (INIS)

    Sadighi, M.; Setayeshi, S.; Salehi, A.A.

    2002-01-01

    This paper presents a new method to solve the problem of finding the best configuration of fuel assemblies in a PWR (Pressurized Water Reactor) core. Finding an optimum solution requires a huge amount of calculations in classical methods. It has been shown that the application of continuous Hop field neural network accompanied by the Simulated Annealing method to this problem not only reduces the volume of the calculations, but also guarantees finding the best solution. In this study flattening of neutron flux inside the reactor core of Brusher NPP is considered as an objective function. The result shows the optimum core configuration which is in agreement with the pattern proposed by the designer

  6. Iodine behaviour in PWR accidents leading to severe core damage

    International Nuclear Information System (INIS)

    Lucas, M.; Devillers, C.; Fermandjian, J.; Manesse, D.

    1982-09-01

    This paper deals with the iodine partition coefficient between the water at the bottom of the reactor building and the atmosphere above it. Molecular iodine is considered as a potential contributor to the airborne activity inside the reactor building. The concentration of molecular iodine in the containment atmosphere will depend, on one hand, upon mechanisms which generate that species and, on the other hand, upon the kinetics of chemical reactions which consume that species. Experiments have therefore been performed on the two following items: - molecular iodine formation through ν radiation from cesium iodide aerosols (droplets) in the reactor containment building, for doses ranging between 1.2 and 8 MRad (12 and 80 kSv), with solutions of various pH's and at different temperatures, - rate of hypoiodous acid disproportionation into iodate and iodide influencing further behavior of molecular iodine

  7. 21 CFR 184.1859 - Invert sugar.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 3 2010-04-01 2009-04-01 true Invert sugar. 184.1859 Section 184.1859 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) FOOD FOR HUMAN... Substances Affirmed as GRAS § 184.1859 Invert sugar. (a) Invert sugar (CAS Reg. No. 8013-17-0) is an aqueous...

  8. Numerical and computational aspects of the coupled three-dimensional core/ plant simulations: organization for economic cooperation and development/ U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-II. 2. TRAB-3D/SMABRE Calculation of the OECD/ NRC PWR MSLB Benchmark

    International Nuclear Information System (INIS)

    Daavittila, A.; Haemaelaeinen, A.; Kyrki-Rajamaki, R.

    2001-01-01

    All three exercises of the OECD/NRC Pressurized Water Reactor (PWR) Main-Steam-Line-Break (MSLB) Benchmark were calculated at VTT Energy. The SMABRE thermal-hydraulics code was used for the first exercise, the plant simulation with point-kinetics neutronics. The second exercise was calculated with the TRAB-3D three-dimensional reactor dynamics code. The third exercise was calculated with the combination TRAB-3D/SMABRE. Both codes have been developed at VTT Energy. The results of all the exercises agree reasonably well with those of the other participants; thus, instead of reporting the results, this paper concentrates on describing the computational aspects of the calculation with the foregoing codes and on some observations of the sensitivity of the results. In the TRAB-3D neutron kinetics, the two-group diffusion equations are solved in homogenized fuel assembly geometry with an efficient two-level nodal method. The point of the two-level iteration scheme is that only one unknown variable per node, the average neutron flux, is calculated during the inner iteration. The nodal flux shapes and cross sections are recalculated only once in the outer iteration loop. The TRAB-3D core model includes also parallel one-dimensional channel hydraulics with detailed fuel models. Advanced implicit time discretization methods are used in all submodels. SMABRE is a fast-running five-equation model completed by a drift-flux model, with a time discretization based on a non-iterative semi-implicit algorithm. For the third exercise of the benchmark, the TMI-1 models of TRAB-3D and SMABRE were coupled. This was the first time these codes were coupled together. However, similar coupling of the HEXTRAN and SMABRE codes has been shown to be stable and efficient, when used in safety analyses of Finnish and foreign VVER-type reactors. The coupling used between the two codes is called a parallel coupling. SMABRE solves the thermal hydraulics both in the cooling circuit and in the core

  9. 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

  10. 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.)

  11. 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

  12. NREL Evaluates Advanced Solar Inverter Performance for Hawaiian Electric

    Science.gov (United States)

    Companies | Energy Systems Integration Facility | NREL NREL Evaluates Advanced Solar Inverter Performance for Hawaiian Electric Companies NREL Evaluates Advanced Solar Inverter Performance for Hawaiian performance and impacts of today's advanced solar inverters, as well as proprietary feedback to the inverter

  13. Development of MHI PWR fuel assembly with high thermal performance

    International Nuclear Information System (INIS)

    Yasushi Makino; Masaya Hoshi; Masaji Mori; Hidetoshi Kido; Kazuo Ikeda

    2005-01-01

    Mitsubishi Heavy Industries, Ltd. (MHI) has been developing a PWR fuel assembly to meet the needs of Japanese fuel market with mainly improving its reliability such as a mechanical strength, a seismic strength and endurance. For burn-up extension of the fuel to 55 GWd/t, MHI has introduced a Zircaloy spacer grid with better neutron economics with retaining the reliability in an operating core. However, for a future power up-rating and a longer cycle operation, a higher thermal performance is required for PWR fuel assembly. To meet the needs of fuel market, MHI has developed an advanced type of Zircaloy spacer grid with a greater DNB performance while retaining the reliability of a fuel and a relatively low pressure drop. For the greater DNB performance, MHI optimized geometrical shape of mixing vane to promote a fluid mixing performance. In this report, higher DNB performance provided by the advanced Zircaloy spacer grid is presented. The results of 3D simulation for the flow behavior in 5 x 5 partial assembly, a mixing test and a water DNB test were compared between the current and the advanced spacer grids. Consequently, it was confirmed that a crossover vane enhanced a fluid mixing and the advanced spacer grid could significantly improve DNB performance compared with the current design of spacer grids. (authors)

  14. Fiber optic components compatibility with the PWR containment radiation field

    International Nuclear Information System (INIS)

    Breuze, G.; Serre, J.

    1990-01-01

    Present and future applications of fiber optics transmission in the nuclear industrial field are emphasized. Nuclear acceptance criteria for relevant electronic equipments in terms of radiation dose rate, integrated dose and required reliability are given. Ambient conditions of PWR containment are especially considered in the present paper. Experimental results of optical fibers and end-components exposed to 60 Co gamma rays are successively shown. Main radiation response characteristics up to 10 4 Gy (with dose rates of about 100 Gy.h -1 ) of both multimodal fiber families (step index and gradient index fibers) are compared. Predominant features of pure silica core fibers are: * an efficient photobleaching with near IR light from LED and LD commonly used in transmission data links, * a radiation hardening reducing induced losses down to 10 dB.km -1 in fine fibers up to date with latest developments. Dose rate effect on induced losses is also outlined for these fibers. Optoelectronic fiber-end components radiation response is good only for special LED (AsGa) and PD (Si). Radiation behavior of complex pigtailed LDM (laser diode + photodiode + Peltier element + thermistor) is not fully acceptable and technological improvements were made. Preliminary results are given. Two applications of fiber links transmitting data in a PWR containment and a hot cell are described. Hardening levels obtained and means required are given

  15. Design and Fabrication of Air-core Inductors for Power Conversion

    DEFF Research Database (Denmark)

    Lê Thanh, Hoà; Mizushima, Io; Tang, Peter Torben

    supply on chip (PwrSoC) [1]. Examples of PwrSoC applications are power adaptors for LED illumination and the “Internet of Things”. We report an air-core MEMS inductor. Our process is scalable and universal for making inductors with versatile geometries e.g. spiral, solenoid, toroid, and advanced...

  16. The analysis of RPV fast neutron flux calculation for PWR with three-dimensional SN method

    International Nuclear Information System (INIS)

    Yang Shouhai; Chen Yixue; Wang Weijin; Shi Shengchun; Lu Daogang

    2011-01-01

    Discrete ordinates (S N ) method is one of the most widely used method for reactor pressure vessel (RPV) design. As the fast development of computer CPU speed and memory capacity and consummation of three-dimensional discrete-ordinates method, it is mature for 3-D S N method to be used to engineering design for nuclear facilities. This work was done specifically for PWR model, with the results of 3-D core neutron transport calculation by 3-D core calculation, 3-D RPV fast neutron flux distribution obtain by 3-D S N method were compared with gained by 1-D and 2-D S N method and the 3-D Monte Carlo (MC) method. In this paper, the application of three-dimensional S N method in calculating RPV fast neutron flux distribution for pressurized water reactor (PWR) is presented and discussed. (authors)

  17. Hydroelastic model of PWR reactor internals SAFRAN 1 - Validation of a vibration calculation method

    International Nuclear Information System (INIS)

    Epstein, A.; Gibert, R.J.; Jeanpierre, F.; Livolant, M.

    1978-01-01

    The SAFRAN 1 test loop consists of an hydroelastic similitude of a 1/8 scale model of a 3 loop P.W.R. Vibrations of the main internals (thermal shield and core barrel) and pressure fluctuations in water thin sections between vessel and internals, and in inlet and outlet pipes, have been measured. The calculation method consists of: an evaluation of the main vibration and acoustic sources owing to the flow (unsteady jet impingement on the core barrel, turbulent flow in a water thin section). A calculation of the internal modal parameters taking into account the inertial effects of fluid (the computer codes AQUAMODE and TRISTANA have been used). A calculation of the acoustic response of the circuit (the computer code VIBRAPHONE has been used). The good agreement between the calculation and the experimental results allows using this method with better security for the prediction of the vibration levels of full scale P.W.R. internals

  18. Fuel management and core design code systems for pressurized water reactor neutronic calculations

    International Nuclear Information System (INIS)

    Ahnert, C.; Arayones, J.M.

    1985-01-01

    A package of connected code systems for the neutronic calculations relevant in fuel management and core design has been developed and applied for validation to the startup tests and first operating cycle of a 900MW (electric) PWR. The package includes the MARIA code system for the modeling of the different types of PWR fuel assemblies, the CARMEN code system for detailed few group diffusion calculations for PWR cores at operating and burnup conditions, and the LOLA code system for core simulation using onegroup nodal theory parameters explicitly calculated from the detailed solutions

  19. Fault-tolerant three-level inverter

    Science.gov (United States)

    Edwards, John; Xu, Longya; Bhargava, Brij B.

    2006-12-05

    A method for driving a neutral point clamped three-level inverter is provided. In one exemplary embodiment, DC current is received at a neutral point-clamped three-level inverter. The inverter has a plurality of nodes including first, second and third output nodes. The inverter also has a plurality of switches. Faults are checked for in the inverter and predetermined switches are automatically activated responsive to a detected fault such that three-phase electrical power is provided at the output nodes.

  20. Development of a computer code for transients simulation in PWR type reactors

    International Nuclear Information System (INIS)

    Alvim, A.C.M.; Botelho, D.A.; Oliveira Barroso, A.C. de

    1981-01-01

    A computer code for the simulation of operacional-transients and accidents in PWR type reactors is being developed at IEN (Instituto de Engenharia Nuclear). Accidents will be considered in which variations in thermohydraulics parameters of fuel and coolant don't cause nucleate boiling in the reactor core, but, otherwise are sufficiently strong to justify a more detailed simulation than that used in linearized models. (E.G.) [pt

  1. Study on evaluating the reactivity worth of the control rods of the PWR 900 MWe

    International Nuclear Information System (INIS)

    Phan Quoc Vuong; Tran Vinh Thanh; Tran Viet Phu

    2015-01-01

    Control rods of a nuclear reactor are divided into two groups: shut down and power control. Reactivity worth of the control rods depends nonlinearly on the rods' compositions and positions where the rods are inserted into the core. Therefore, calculation of control rod worth is of high important. In this study, we calculated the reactivity worth of the power control rod bank of the Mitsubishi PWR 900 MWe. The results are integral and differential worth calibration of the control rods. (author)

  2. The computer program ELCOM in the planning and structural analysis of PWR fuel elements: an example

    International Nuclear Information System (INIS)

    Silva Macedo, L.V. da

    1990-01-01

    Is's presented some results obtained with the ELCOM computer code, such as deflections, moments and natural frequencies, used in the design and structural analysis of PWR fuels assemblies. It's studied the behavior of these results varying the number of spacer grids, the rigidity of the joint between the fuel pin and the spacer grid, and the fuel assembly's boundary condition, considered in the analysis, in it's mounting into the core (if clamped-clamped, clamped-hinged or hinged-hinged). (author)

  3. Integrated Inverter And Battery Charger

    Science.gov (United States)

    Rippel, Wally E.

    1988-01-01

    Circuit combines functions of dc-to-ac inversion (for driving ac motor in battery-powered vehicle) and ac-to-dc conversion (for charging battery from ac line when vehicle not in use). Automatically adapts to either mode. Design of integrated inverter/charger eliminates need for duplicate components, saves space, reduces weight and cost of vehicle. Advantages in other applications : load-leveling systems, standby ac power systems, and uninterruptible power supplies.

  4. Effective switching frequency multiplier inverter

    Science.gov (United States)

    Su, Gui-Jia [Oak Ridge, TN; Peng, Fang Z [Okemos, MI

    2007-08-07

    A switching frequency multiplier inverter for low inductance machines that uses parallel connection of switches and each switch is independently controlled according to a pulse width modulation scheme. The effective switching frequency is multiplied by the number of switches connected in parallel while each individual switch operates within its limit of switching frequency. This technique can also be used for other power converters such as DC/DC, AC/DC converters.

  5. Control rod effects with plutonium recycle in a PWR

    International Nuclear Information System (INIS)

    Nash, G.; Muehl, G.J.; Gibson, I.H.

    1979-03-01

    A study has been made on a PWR loaded partly and wholly with plutonium to determine the changes in shutdown margin compared with an enriched uranium core. Lattice calculations are used to generate cell constants for core calculations. Three fuel loadings were considered, all uranium, 30% (approximately) of the assemblies plutonium in natural uranium, and all plutonium. The equilibrium fuel management schemes adopted in each case are based on the standard three cycle equal size batch scheme. Detailed calculations of power and irradiation distributions through the cycles have been carried out to provide a starting point for the control rod worth and requirement calculations. Control rod worths are reduced in a plutonium core because of the harder spectrum and higher fuel absorption cross sections. Furthermore, the control rod requirements for shutdown increase because of the increase in fuel and moderator temperature coefficients. This results in a reduction in shutdown margin. The magnitude of these changes is fully analysed in the report. The significance of these reductions depends on the detail of the safety argument but reductions of these sizes are unlikely to be acceptable. The data provided in this report could be used to give a first estimate of the plutonium loading acceptable given the safety assessment of the normal uranium core. (U.K.)

  6. 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.

  7. 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.)

  8. Source terms associated with two severe accident sequences in a 900 MWe PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.; Berthion, Y.; Lhiaubet, G.; Lucas, M.

    1983-12-01

    Hypothetical accidents taken into account in PWR risk assessment result in fission product release from the fuel, transfer through the primary circuit, transfer into the reactor containment building (RCB) and finally release to the environment. The objective of this paper is to define the characteristics of the source term (noble gases, particles and volatile iodine forms) released from the reactor containment building during two dominant core-melt accident sequences: S 2 CD and TLB according to the ''Reactor Safety Study'' terminology. The reactor chosen for this study is a French 900 MWe PWR unit. The reactor building is a prestressed concrete containment with an internal liner. The first core-melt accident sequence is a 2-break loss-of-coolant accident on the cold leg, with failure of both system and the containment spray system. The second one is a transient initiated by a loss of offsite and onsite power supply and auxiliary feedwater system. These two sequences have been chosen because they are representative of risk dominant scenarios. Source terms associated with hypothetical core-melt accidents S 2 CD and TLB in a French PWR -900 MWe- have been performed using French computer codes (in particular, JERICHO Code for containment response analysis and AEROSOLS/31 for aerosol behavior in the containment)

  9. 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)

  10. 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

  11. Simulation of fission products behavior in severe accidents for advanced passive PWR

    International Nuclear Information System (INIS)

    Tong, L.L.; Huang, G.F.; Cao, X.W.

    2015-01-01

    Highlights: • A fission product analysis model based on thermal hydraulic module is developed. • An assessment method for fission product release and transport is constructed. • Fission products behavior during three modes of containment response is investigated. • Source term results for the three modes of containment response are obtained. - Abstract: Fission product behavior for common Pressurized Water Reactor (PWR) has been studied for many years, and some analytical tools have developed. However, studies specifically on the behavior of fission products related to advanced passive PWR is scarce. In the current study, design characteristics of advanced passive PWR influencing fission product behavior are investigated. An integrated fission products analysis model based on a thermal hydraulic module is developed, and the assessment method for fission products release and transport for advanced passive PWR is constructed. Three modes of containment response are simulated, including intact containment, containment bypass and containment overpressure failure. Fission products release from the core and corium, fission products transport and deposition in the Reactor Coolant System (RCS), fission products transport and deposition in the containment considering fission products retention in the in-containment refueling water storage tank (IRWST) and in the secondary side of steam generators (SGs) are simulated. Source term results of intact containment, containment bypass and containment overpressure failure are obtained, which can be utilized to evaluate the radiological consequences

  12. CFD simulation of a four-loop PWR at asymmetric operation conditions

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, Jian-Ping; Yan, Li-Ming; Li, Feng-Chen, E-mail: lifch@hit.edu.cn

    2016-04-15

    Highlights: • A CFD numerical simulation procedure was established for simulating RPV of VVER-1000. • The established CFD approach was validated by comparing with available data. • Thermal hydraulic characteristics under asymmetric operation condition were investigated. • Apparent influences of the shutdown loop on its neighboring loops were obtained. - Abstract: The pressurized water reactor (PWR) with multiple loops may have abnormal working conditions with coolant pumps out of running in some loops. In this paper, a computational fluid dynamics (CFD) numerical study of the four-loop VVER-1000 PWR pressure vessel model was presented. Numerical simulations of the thermohydrodynamic characteristics in the pressure vessel were carried out at different inlet conditions with four and three loops running, respectively. At normal stead-state condition (four-loop running), different parameters were obtained for the full fluid domain, including pressure losses across different parts, pressure, velocity and temperature distributions in the reactor pressure vessel (RPV) and mass flow distribution of the coolant at the inlet of reactor core. The obtained results for pressure losses matched with the experimental reference values of the VVER-1000 PWR at Tianwan nuclear power plant (NPP). For most fuel assemblies (FAs), the inlet flow rates presented a symmetrical distribution about the center under full-loop operation conditions, which accorded with the practical distribution. These results indicate that it is now possible to study the dynamic transition process between different asymmetric operation conditions in a multi-loop PWR using the established CFD method.

  13. Evaluation of fire probabilistic safety assessment for a PWR plant

    International Nuclear Information System (INIS)

    Wu, C.H.; Lin, T.J.; Kao, T.M.

    2001-01-01

    The internal fire analysis of the level 1 power operation probability safety assessment (PSA) for Maanshan (PWR) Nuclear Power Plant (MNPP) was updated. The fire analysis adopted a scenario-based PSA approach to systematically evaluate fire and smoke hazards and their associated risk impact to MNPP. The result shows that the core damage frequency (CDF) due to fire is about six times lower than the previous one analyzed by the Atomic Energy Council (AEC), Republic of China in 1987. The plant model was modified to reflect the impact of human events and recovery actions during fire. Many tabulated EXCEL spread-sheets were used for evaluation of the fire risk. The fire-induced CDF for MNPP is found to be 2.1 E-6 per year in this study. The relative results of the fire analysis will provide the bases for further risk-informed fire protection evaluation in the near future. (author)

  14. Shutdown Chemistry Process Development for PWR Primary System

    Energy Technology Data Exchange (ETDEWEB)

    Sung, K.B. [Korea Electric Power Research Institute, Taejeon (Korea, Republic of)

    1997-12-31

    This study report presents the shutdown chemistry of PWR primary system to reduce and remove the radioactive corrosion products which were deposited on the nuclear fuel rods surface and the outside of core like steam generator channel head, RCS pipings etc. The major research results are the follows ; the deposition radioactive mechanism of corrosion products, the radiochemical composition, the condition of coolant chemistry to promote the dissolution of radioactive cobalt and nickel ferrite, the control method of dissolved hydrogen concentration in the coolant by the mechanical and chemical methods. The another part of study is to investigate the removal characteristics of corrosion product ions and particles by the demineralization system to suggest the method which the system could be operate effectively in shut-down purification period. (author). 19 refs., 25 figs., 48 tabs.

  15. Reduced scale PWR passive safety system designing by genetic algorithms

    International Nuclear Information System (INIS)

    Cunha, Joao J. da; Alvim, Antonio Carlos M.; Lapa, Celso Marcelo Franklin

    2007-01-01

    This paper presents the concept of 'Design by Genetic Algorithms (DbyGA)', applied to a new reduced scale system problem. The design problem of a passive thermal-hydraulic safety system, considering dimensional and operational constraints, has been solved. Taking into account the passive safety characteristics of the last nuclear reactor generation, a PWR core under natural circulation is used in order to demonstrate the methodology applicability. The results revealed that some solutions (reduced scale system DbyGA) are capable of reproducing, both accurately and simultaneously, much of the physical phenomena that occur in real scale and operating conditions. However, some aspects, revealed by studies of cases, pointed important possibilities to DbyGA methodological performance improvement

  16. PWR cold-leg small break loca with faulty HPI

    International Nuclear Information System (INIS)

    Kumamaru, H.; Kukita, Y.

    1991-01-01

    The ROSA-IV Large Scale Test Facility (LSTF) is a 1/48 volumetrically-scaled model of a pressurized water reactor (PWR). At the LSTF are performed cold-leg small-break loss-of-coolant accident (LOCA) tests with faulty high pressure injection (HPI) system for break areas from 0.5% to 10% and an intentional primary system depressurization test following a small-break LOCA test. A simple prediction model is proposed for prediction of times of major events. Test data and calculations show that intentional primary system depressurization with use of the pressurizer power-operated relief valves (PORVs) is effective for break areas of approximately 0.5% or less, is unnecessary for breaks of 5% or more, and is insufficient for intermediate break areas to maintain adequate core cooling. (author)

  17. Fuel cycle optimization in PWR'S

    International Nuclear Information System (INIS)

    Castro Lobo, P.D. de; Amorim, E.S. do.

    1979-08-01

    Neutronics aspects of a reactor core throughout its cycle were investigated in a search for increasing in-core utilization of the residual fissile isotopes content in the cycle discharged disposal. The effects due to design modifications introduced at burnup levels near the end-of-cycle, in an equilibrium cycle condition, have indicated the possibility of a better in-core utilization of the residual fissile isotopes existing in the cycle discharged disposal. The potential benefits are significant to warranty an examination of the mechanical and thermal hydraulic involved. At convenient burnup levels, change in H 2 O/UO 2 volume ratio were introduced allowing an intense depletion of the residual fissile isotopes existing in assemblies with high exposures levels. (Author) [pt

  18. 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)

  19. 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)

  20. The design of a compact integral medium size PWR

    International Nuclear Information System (INIS)

    Shirvan, Koroush; Hejzlar, Pavel; Kazimi, Mujid S.

    2012-01-01

    Highlights: ► We model the IRIS reactor in RELAP5 and VIPRE codes. ► We use Printed Circuit Heat Exchangers and internally and externally cooled fuel pins in IRIS. ► We increase the IRIS power by 50% and demonstrate adequate safety performance. ► We show significant potential gain in economics for any integral PWR reactor design. - Abstract: Integral reactors have been proposed in recent years as a means to eliminate loss of coolant events, and reduce the number of large vessels of a nuclear power plant. In this paper the focus on how to further increase the power that can be derived from a given vessel volume. The example is applied to the International Reactor Innovative and Secure (IRIS), a medium size, light water reactor rated at 1000 MWt. The IRIS is an integral design containing all pumps and steam generators along with a traditional PWR core inside the reactor vessel. IRIS was designed with 8 Once-Through Helically Coiled Steam Generators (OTHSG), located above the core, in an annular region between the riser and the pressure vessel wall. This work examines ideas to increase its power output in the same vessel size while maintaining or improving the safety margins. The combination of Printed Circuit Heat Exchangers (PCHE) and Internally and EXternally cooled Annular Fuel (IXAF) is proposed to implement such improvement in otherwise the reference IRIS design. Safety implications of such steam generator and fuel design changes for the same reactor size are examined, under both steady state and transients, using the RELAP5 and VIPRE codes. It is found that the IRIS reactor power can be increased by 50% by using the PCHE and IXAF. The proposed design is found to be less expensive per unit electric power produced, these improvements and analyses can be applied to any integral reactor design.

  1. PWR control rod ejection analysis with the numerical nuclear reactor

    International Nuclear Information System (INIS)

    Hursin, M.; Kochunas, B.; Downar, T. J.

    2008-01-01

    During the past several years, a comprehensive high fidelity reactor LWR core modeling capability has been developed and is referred to as the Numerical Nuclear Reactor (NNR). The NNR achieves high fidelity by integrating whole core neutron transport solution and ultra fine mesh computational fluid dynamics/heat transfer solution. The work described in this paper is a preliminary demonstration of the ability of NNR to provide a detailed intra pin power distribution during a control rod ejection accident. The motivation of the work is to quantify the impact on the fuel performance calculation of a more physically accurate representation of the power distribution within the fuel rod during the transient. The paper addresses first, the validation of the transient capability of the neutronic module of the NNR code system, DeCART. For this purpose, a 'mini core' problem consisting of a 3x3 array of typical PWR fuel assemblies is considered. The initial state of the 'mini core' is hot zero power with a control rod partially inserted into the central assembly which is fresh fuel and is adjacent to once and twice burned fuel representative of a realistic PWR arrangement. The thermal hydraulic feedbacks are provided by a simplified fluids and heat conduction solver consistent for both PARCS and DeCART. The control rod is ejected from the central assembly and the transient calculation is performed with DeCART and compared with the results of the U.S. NRC core simulation code PARCS. Because the pin power reconstruction in PARCS is based on steady state intra assembly pin power distributions which do not account for thermal feedback during the transient and which do not take into account neutron leakage from neighboring assemblies during the transient, there are some small differences in the PARCS and DeCART pin power prediction. Intra pin power density information obtained with DeCART represents new information not available with previous generation of methods. The paper then

  2. Containment loading during severe core damage accidents

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.; Cenerino, C.; Berthion, Y.; Carvallo, G.

    1984-11-01

    The objective of the article is to study the influence of the state of the reactor cavity (dry or flooded) and of the corium coolability on the thermal-hydraulics in the containment in the case of an accident sequence involving core melting and subsequent containment basemat erosion, in a 900 MWe PWR unit. Calculations are performed by using the JERICHO thermal hydraulics code

  3. Analysis of the return to power scenario following a LBLOCA in a PWR

    Energy Technology Data Exchange (ETDEWEB)

    Macian, R.; Tyler, T.N.; Mahaffy, J.H. [Pennsylvania State Univ., University Park, PA (United States)

    1995-09-01

    The risk of reactivity accidents has been considered an important safety issue since the beginning of the nuclear power industry. In particular, several events leading to such scenarios for PWR`s have been recognized and studied to assess the potential risk of fuel damage. The present paper analyzes one such event: the possible return to power during the reflooding phase following a LBLOCA. TRAC-PF1/MOD2 coupled with a three-dimensional neutronic model of the core based on the Nodal Expansion Method (NEM) was used to perform the analysis. The system computer model contains a detailed representation of a complete typical 4-loop PWR. Thus, the simulation can follow complex system interactions during reflooding, which may influence the neutronics feedback in the core. Analyses were made with core models bases on cross sections generated by LEOPARD. A standard and a potentially more limiting case, with increased pressurizer and accumulator inventories, were run. In both simulations, the reactor reaches a stable state after the reflooding is completed. The lower core region, filled with cold water, generates enough power to boil part of the incoming liquid, thus preventing the core average liquid fraction from reaching a value high enough to cause a return to power. At the same time, the mass flow rate through the core is adequate to maintain the rod temperature well below the fuel damage limit.

  4. Control rod ejection accident analysis for a PWR with thorium fuel loading

    Energy Technology Data Exchange (ETDEWEB)

    Da Cruz, D.F. [Nuclear Research and Consultancy Group NRG, Westerduinweg 3, P.O. Box 25, 1755 ZG Petten (Netherlands)

    2010-07-01

    This paper presents the results of 3-D transient analysis of a pressurized water reactor (PWR) core loaded with 100% Th-Pu MOX fuel assemblies. The aim of this study is to evaluate the safety impact of applying a full loading of this innovative fuel in PWRs of the current generation. A reactivity insertion accident scenario has been simulated using the reactor core analysis code PANTHER, used in conjunction with the lattice code WIMS. A single control rod assembly, with the highest reactivity worth, has been considered to be ejected from the core within 100 milliseconds, which may occur due to failure of the casing of the control rod driver mechanism. Analysis at both hot full power and hot zero power reactor states have been taken into account. The results were compared with those obtained for a representative PWR fuelled with UO{sub 2} fuel assemblies. In general the results obtained for both cores were comparable, with some differences associated mainly to the harder neutron spectrum observed for the Th-Pu MOX core, and to some specific core design features. The study has been performed as part of the LWR-DEPUTY project of the EURATOM 6. Framework Programme, where several aspects of novel fuels are being investigated for deep burning of plutonium in existing nuclear power plants. (authors)

  5. Understanding and managing the effects of battery charger and inverter aging

    International Nuclear Information System (INIS)

    Gunther, W.; Aggarwal, S.

    1992-01-01

    An aging assessment of battery chargers and inverters was conducted under the auspices of the NRC's Nuclear Plant Aging Research (NPAR) Program. The intentions of this program are to resolve issues related to the aging and service wear of equipment and systems at operating reactor facilities and to assess their impact on safety. Inverters and battery chargers are used in nuclear power plants to perform significant functions related to plant safety and availability. The specific impact of a battery charger or inverter failure varies with plant configuration. Operating experience data have demonstrated that reactor trips, safety injection system actuations, and inoperable emergency core cooling systems have resulted from inverter failures; and dc bus degradation leading to diesel generator inoperability or loss of control room annunication and indication have resulted from battery and battery charger failures. For the battery charger and inverter, the aging and service wear of subcomponents have contributed significantly to equipment failures. This paper summarizes the data and then describes methods that can be used to detect battery charger and inverter degradation prior to failure, as well as methods to minimize the failure effects. In both cases, the managing of battery charger and inverter aging is emphasized. 5 refs

  6. Multilevel Inverter by Cascading Industrial VSI

    DEFF Research Database (Denmark)

    Teodorescu, Remus; Blaabjerg, Frede; Pedersen, John Kim

    2002-01-01

    In this paper the modularity concept applied to medium-voltage adjustable speed drives is addressed. First, the single-phase cascaded voltage-source inverter that uses series connection of IGBT H-bridge modules with isolated dc-buses is presented. Next, a novel three-phase cascaded voltage......-source inverter that uses three IGBT triphase inverter modules along with an output transformer to obtain a 3 p.u. multilevel output voltage is introduced. The system yields in high-quality multistep voltage with up to 4 levels and low dv/dt, balanced operation of the inverter modules, each supplying a third...... of the motor rated kVA. The concept of using cascaded inverters is further extended to a new modular motor-modular inverter system where the motor winding connections are reconnected into several three-phase groups, either six-lead or 12-lead connection according to the voltage level, each powered...

  7. Fluid-structure interactions in PWR vessels during blowdown

    International Nuclear Information System (INIS)

    Schumann, U.; Enderle, G.; Katz, F.; Ludwig, A.; Moesinger, H.; Schlechtendahl, E.G.

    1979-01-01

    For analysis of blowdown loadings and dynamic response of PWR vessel internals several computer codes have been developed at Karlsruhe. The goal is to provide advanced codes which permit a 'best estimate' analysis of the deformations and stresses of the internal structures, in particular the core barrel, such that the safety margins can be evaluated. The stresses reach their maxima during the initial subcooled period of the blowdown in which two-phase phenomena are important in the blowdown pipe only. In this period, the computed results with and without fluid-structural interactions show that the coupling between the water in the downcomer and the rather thin elastic core barrel is of dominant importance. Without coupling the core barrel oscillates with much higher frequencies than with coupling. The amplitudes and stresses are about twice as large initially. Later, the decoupled analysis can result in a meaningless overestimation of the structural response. By comparison of computations for incompressible and for compressible fluid with and without coupling we have found that a correct treatment of the fluid-structure coupling is more important than the description of pressure waves. (orig.)

  8. PWR loading pattern optimization using Harmony Search algorithm

    International Nuclear Information System (INIS)

    Poursalehi, N.; Zolfaghari, A.; Minuchehr, A.

    2013-01-01

    Highlights: ► Numerical results reveal that the HS method is reliable. ► The great advantage of HS is significant gain in computational cost. ► On the average, the final band width of search fitness values is narrow. ► Our experiments show that the search approaches the optimal value fast. - Abstract: In this paper a core reloading technique using Harmony Search, HS, is presented in the context of finding an optimal configuration of fuel assemblies, FA, in pressurized water reactors. To implement and evaluate the proposed technique a Harmony Search along Nodal Expansion Code for 2-D geometry, HSNEC2D, is developed to obtain nearly optimal arrangement of fuel assemblies in PWR cores. This code consists of two sections including Harmony Search algorithm and Nodal Expansion modules using fourth degree flux expansion which solves two dimensional-multi group diffusion equations with one node per fuel assembly. Two optimization test problems are investigated to demonstrate the HS algorithm capability in converging to near optimal loading pattern in the fuel management field and other subjects. Results, convergence rate and reliability of the method are quite promising and show the HS algorithm performs very well and is comparable to other competitive algorithms such as Genetic Algorithm and Particle Swarm Intelligence. Furthermore, implementation of nodal expansion technique along HS causes considerable reduction of computational time to process and analysis optimization in the core fuel management problems

  9. 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

  10. 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

  11. 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

  12. 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)

  13. 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)

  14. 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

  15. 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)

  16. Development of computational methods to describe the mechanical behavior of PWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Wanninger, Andreas; Seidl, Marcus; Macian-Juan, Rafael [Technische Univ. Muenchen, Garching (Germany). Dept. of Nuclear Engineering

    2016-10-15

    To investigate the static mechanical response of PWR fuel assemblies (FAs) in the reactor core, a structural FA model is being developed using the FEM code ANSYS Mechanical. To assess the capabilities of the model, lateral deflection tests are performed for a reference FA. For this purpose we distinguish between two environments, in-laboratory and in-reactor for different burn-ups. The results are in qualitative agreement with experimental tests and show the stiffness decrease of the FAs during irradiation in the reactor core.

  17. The probability of containment failure by steam explosion in a PWR

    International Nuclear Information System (INIS)

    Briggs, A.J.

    1983-12-01

    The study of the risk associated with operation of a PWR includes assessment of severe accidents in which a combination of faults results in melting of the core. Probabilistic methods are used in such assessment, hence it is necessary to estimate the probability of key events. One such event is the occurrence of a large steam explosion when molten core debris slumps into the base of the reactor vessel. This report considers recent information, and recommends an upper limit to the range of probability values for containment failure by steam explosion for risk assessment for a plant such as the proposed Sizewell B station. (U.K.)

  18. Special points of view about development and construction of a PWR

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1977-01-01

    1.0) The reactor core and its components, 1.1) design of the fuel assemblies, 1.2) incore instrumentation, 2.0) reactor pressure vessel with internals, 3.0) components of the reactor coolant loops, 3.1) steam generator, 3.2) pressurizer, 3.3) pressurizer relief tank, 3.4) reactor coolant pumps, 4.0) instrumentation and control of a PWR, 4.1) ex-core measuring system, 4.2) reactor protection system, 4.3) control systems, 4.4) radiation monitoring. (orig.) [de

  19. Improvements to PWR type reactors

    International Nuclear Information System (INIS)

    Ailloud, Jean; Monteil, Marcel.

    1978-01-01

    Improvements to pressurized water nuclear reactors are described, where the core coolant, called primary fluid, flows under the effect of a circulating pump in a primary loop between a steam generator and a pressure vessel containing the reactor core. The steam generator includes a bundle of tubes through which flows the primary fluid which exchanges calories with a secondary fluid, generally water, entering the generator as a liquid and issuing from it as steam. After expansion in turbines and recovery in a condenser, this steam is returned to the inside of the generator. Each primary fluid circulating pump is powered by a back-pressure turbine located in parallel with the high pressure section of the main turbine and hence fed with steam taken directly from the steam generator or the main steam pipe outside it [fr

  20. Multiplexing and data processing of in-core signals

    International Nuclear Information System (INIS)

    Meyer, M.

    1983-01-01

    The application of multiplexing and signal processing techniques used for reactor operation and utilisation of data from the in-core instrumentation system is described. After a brief recall about in-core instrumentation, the aims, the advantages of multiplexing are presented. One of the aims of this realization is to show the compatibity between the technologies used with a PWR environment [fr

  1. Experimental research progress on passive safety systems of Chinese advanced PWR

    International Nuclear Information System (INIS)

    Xiao Zejun; Zhuo Wenbin; Zheng Hua; Chen Bingde; Zong Guifang; Jia Dounan

    2003-01-01

    TMI and Chernobyl accidents, having pronounced impact on nuclear industries, triggered the governments as well as interested institutions to devote much attention to the safety of nuclear power plant and public's requirements on nuclear power plant safety were also going to be stricter and stricter. It is obvious that safety level of an ordinary light water reactor is no longer satisfactory to these requirements. Recently, the safety authorities have recommended the implementation of passive system to improve the safety of nuclear reactors. Passive safety system is one of the main differences between Chinese advanced PWR and other conventional PWR. The working principle of passive safety system is to utilize the gravity, natural convection (natural circulation) and stored energy to implement the system's safety function. Reactors with passive safety systems are not only safer, but also more economical. The passive safety system of Chinese advanced PWR is composed of three independent systems, i.e. passive containment cooling system, passive residual heat removal system and passive core makeup tank injection system. This paper is a summary of experimental research progress on passive containment cooling system, passive residual heat removal system and passive core makeup tank injection system

  2. Development of a thermal-hydraulic code for reflood analysis in a PWR experimental loop

    International Nuclear Information System (INIS)

    Alves, Sabrina P.; Mesquita, Amir Z.; Rezende, Hugo C.; Palma, Daniel A.P.

    2017-01-01

    A process of fundamental importance in the event of Loss of Coolant Accident (LOCA) in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Center (CDTN) has been developing since the 70’s programs to allow Brazil to become independent in the field of reactor safety analysis. To that end, in the 80’s was designed, assembled and commissioned one Rewetting Test Facility (ITR in Portuguese). This facility aims to investigate the phenomena involved in the thermal hydraulic reflood phase of a Loss of Coolant Accident in a PWR nuclear reactor. This work aim is the analysis of physical and mathematical models governing the rewetting phenomenon, and the development a thermo-hydraulic simulation code of a representative experimental circuit of the PWR reactors core cooling channels. It was possible to elaborate and develop a code called REWET. The results obtained with REWET were compared with the experimental results of the ITR, and with the results of the Hydroflut code, that was the old program previously used. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the front for two typical tests using the two codes calculation, and experimental results. The result simulated by REWET code for the rewetting time also came closer to the experimental results more than those calculated by Hydroflut code. (author)

  3. Development of a thermal-hydraulic code for reflood analysis in a PWR experimental loop

    Energy Technology Data Exchange (ETDEWEB)

    Alves, Sabrina P.; Mesquita, Amir Z.; Rezende, Hugo C., E-mail: sabrinapral@gmail.com, E-mail: amir@cdtn.brm, E-mail: hcr@cdtn.br, E-mail: hcr@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissão Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    A process of fundamental importance in the event of Loss of Coolant Accident (LOCA) in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Center (CDTN) has been developing since the 70’s programs to allow Brazil to become independent in the field of reactor safety analysis. To that end, in the 80’s was designed, assembled and commissioned one Rewetting Test Facility (ITR in Portuguese). This facility aims to investigate the phenomena involved in the thermal hydraulic reflood phase of a Loss of Coolant Accident in a PWR nuclear reactor. This work aim is the analysis of physical and mathematical models governing the rewetting phenomenon, and the development a thermo-hydraulic simulation code of a representative experimental circuit of the PWR reactors core cooling channels. It was possible to elaborate and develop a code called REWET. The results obtained with REWET were compared with the experimental results of the ITR, and with the results of the Hydroflut code, that was the old program previously used. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the front for two typical tests using the two codes calculation, and experimental results. The result simulated by REWET code for the rewetting time also came closer to the experimental results more than those calculated by Hydroflut code. (author)

  4. Loss-of-Fluid Test findings in pressurized water reactor core's thermal-hydraulic behavior

    International Nuclear Information System (INIS)

    Russell, M.

    1983-01-01

    This paper summarizes the pressurized water reactor (PWR) core's thermal-hydraulic behavior findings from experiments performed at the Loss-of-Fluid Test (LOFT) Facility at the Idaho National Engineering Laboratory. The potential impact of these findings on the safety and economics of PWR's generation of electricity is also discussed. Reviews of eight important findings in the core's physical behavior and in experimental methods are presented with supporting evidence

  5. Comparison of Serpent and HELIOS-2 as applied for the PWR few-group cross section generation

    International Nuclear Information System (INIS)

    Fridman, E.; Leppaenen, J.; Wemple, C.

    2013-01-01

    This paper discusses recent modifications to the Serpent Monte Carlo code methodology and related to the calculation of few-group diffusion coefficients and reflector discontinuity factors The new methods were assessed in the following manner. First, few-group homogenized cross sections calculated by Serpent for a reference PWR core were compared with those generated 1 commercial deterministic lattice transport code HELIOS-2. Second, Serpent and HELIOS-2 fe group cross section sets were later employed by nodal diffusion code DYN3D for the modeling the reference PWR core. Finally, the nodal diffusion results obtained using the both cross section sets were compared with the full core Serpent Monte Carlo solution. The test calculations show that Serpent can calculate the parameters required for nodal analyses similar to conventional deterministic lattice codes. (authors)

  6. Sensitivity of risk parameters to human errors in reactor safety study for a PWR

    International Nuclear Information System (INIS)

    Samanta, P.K.; Hall, R.E.; Swoboda, A.L.

    1981-01-01

    Sensitivities of the risk parameters, emergency safety system unavailabilities, accident sequence probabilities, release category probabilities and core melt probability were investigated for changes in the human error rates within the general methodological framework of the Reactor Safety Study (RSS) for a Pressurized Water Reactor (PWR). Impact of individual human errors were assessed both in terms of their structural importance to core melt and reliability importance on core melt probability. The Human Error Sensitivity Assessment of a PWR (HESAP) computer code was written for the purpose of this study. The code employed point estimate approach and ignored the smoothing technique applied in RSS. It computed the point estimates for the system unavailabilities from the median values of the component failure rates and proceeded in terms of point values to obtain the point estimates for the accident sequence probabilities, core melt probability, and release category probabilities. The sensitivity measure used was the ratio of the top event probability before and after the perturbation of the constituent events. Core melt probability per reactor year showed significant increase with the increase in the human error rates, but did not show similar decrease with the decrease in the human error rates due to the dominance of the hardware failures. When the Minimum Human Error Rate (M.H.E.R.) used is increased to 10 -3 , the base case human error rates start sensitivity to human errors. This effort now allows the evaluation of new error rate data along with proposed changes in the man machine interface

  7. Lightweight submersed 'Walking' NDE manipulators for PWR and BWR vessel weld inspection

    International Nuclear Information System (INIS)

    Saernmark, Ivan; Lenz, Herbert

    2008-01-01

    Three new manipulators developed by WesDyne TRC in Sweden have under the year 2007 performed three very successful inspections in the PWR reactor Ringhals 3 and the BWR reactors Ringhals 1 and Oskarshamn 1. The manipulator systems can be used to perform inspection of circumferential and vertical welds on the reactor pressure vessel, the core shroud, core shroud support in BWR reactors or vessel and core barrel welds in PWR reactors. Most other flat or curved surfaces can be inspected using the new concept through relatively simple mechanical reconfigurations of system modules. The first inspection was performed on the R3 PWR core barrel in June 2007 with a very good result. This Manipulator is designed for access in very narrow gaps and for the type of core barrels with a shield covering the whole area of the perimeter. The manipulator is attached to the inspection area by means of a new unique suction cup system. The current manipulators consist of a curved horizontal beam, with radius similar to the reactor vessel, and a straight vertical beam, forming a T-shaped structure. By alternating the application of suction cup pairs on the horizontal beam and the vertical beam and by driving the scanning motors, the manipulator performs an incremental translational movement upwards/downwards or from side to side. The principles of this system give a well defined and stable platform for global and local positioning accuracy. A combination of advanced sensor solutions provides accurate position information in the absence of other physical reference objects. The system is controlled by the new WesDyne TRC Motor Control Panel and software, the MCP is specifically designed for remote control of submersed manipulators using techniques for cable reduction

  8. Lightweight submersed 'Walking' NDE manipulators for PWR and BWR vessel weld inspection

    Energy Technology Data Exchange (ETDEWEB)

    Saernmark, Ivan; Lenz, Herbert [WesDyne TRC AB, Stockholm (Sweden)

    2008-04-15

    Three new manipulators developed by WesDyne TRC in Sweden have under the year 2007 performed three very successful inspections in the PWR reactor Ringhals 3 and the BWR reactors Ringhals 1 and Oskarshamn 1. The manipulator systems can be used to perform inspection of circumferential and vertical welds on the reactor pressure vessel, the core shroud, core shroud support in BWR reactors or vessel and core barrel welds in PWR reactors. Most other flat or curved surfaces can be inspected using the new concept through relatively simple mechanical reconfigurations of system modules. The first inspection was performed on the R3 PWR core barrel in June 2007 with a very good result. This Manipulator is designed for access in very narrow gaps and for the type of core barrels with a shield covering the whole area of the perimeter. The manipulator is attached to the inspection area by means of a new unique suction cup system. The current manipulators consist of a curved horizontal beam, with radius similar to the reactor vessel, and a straight vertical beam, forming a T-shaped structure. By alternating the application of suction cup pairs on the horizontal beam and the vertical beam and by driving the scanning motors, the manipulator performs an incremental translational movement upwards/downwards or from side to side. The principles of this system give a well defined and stable platform for global and local positioning accuracy. A combination of advanced sensor solutions provides accurate position information in the absence of other physical reference objects. The system is controlled by the new WesDyne TRC Motor Control Panel and software, the MCP is specifically designed for remote control of submersed manipulators using techniques for cable reduction.

  9. Three-Phase Cascaded Multilevel Inverter Using Power Cells With Two Inverter Legs in Series

    NARCIS (Netherlands)

    Waltrich, G.; Barbi, I.

    2010-01-01

    In this paper, a modular three-phase multilevel inverter specially suited for electrical drive applications is proposed. Unlike the cascaded H-bridge inverter, this topology is based on power cells connected in cascade using two inverter legs in series. A detailed analysis of the structure and the

  10. DOMPAC dosimetry experiment. Neutronic simulation of the thickness of a PWR pressure vessel. Irradiation damages

    International Nuclear Information System (INIS)

    Alberman, A.; Faure, M.; Thierry, M.; Hoclet, O.; Le Dieu de Ville, A.; Nimal, J.C.; Soulat, P.

    1979-01-01

    For suitable extrapolation of irradiated PWR ferritic steel results, proper irradiation of the pressure vessel has been 'simulated' in test reactor. For this purpose, a huge steel block (20 cm in depth) was loaded with Saclay's graphite (GAMIN) and tungsten damage detectors. Core-block water gap was optimized through spectrum indexes method, by ANISN and SABINE codes so that spectrum in 1/4 thickness matches with ANISN computations for PWR Fessenheim 1. A good experimental agreement is found with calculated dpa damage gradient. 3D Monte Carlo computation (TRIPOLI), was performed on the DOMPAC device, and spectrum indexes evolution was found consistent with experimental results. Surveillance rigs behind a 'thermal shield' were also simulated, including damage and activation monitors. Dosimetry results give an order of magnitude of accuracies involved in projecting steel sample embrittlement to the pressure vessel [fr

  11. Conceptual design study of small long-life PWR based on thorium cycle fuel

    International Nuclear Information System (INIS)

    Subkhi, M. Nurul; Su'ud, Zaki; Waris, Abdul; Permana, Sidik

    2014-01-01

    A neutronic performance of small long-life Pressurized Water Reactor (PWR) using thorium cycle based fuel has been investigated. Thorium cycle which has higher conversion ratio in thermal region compared to uranium cycle produce some significant of 233 U during burn up time. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.3, while the multi-energy-group diffusion calculations were optimized in whole core cylindrical two-dimension R-Z geometry by SRAC-CITATION. this study would be introduced thorium nitride fuel system which ZIRLO is the cladding material. The optimization of 350 MWt small long life PWR result small excess reactivity and reduced power peaking during its operation

  12. Power ramp testing method for PWR fuel rod at research reactor

    International Nuclear Information System (INIS)

    Zhou Yidong; Zhang Peisheng; Zhang Aimin; Gao Yongguang; Wang Huarong

    2003-01-01

    A tentative power ramp test for short PWR fuel rod has been conducted at the Heavy Water Research Reactor (HWRR) in China Institute of Atomic Energy (CIAE). The test fuel rod was cooled by the circulating water in the test loop. The power ramp was realized by moving solid neutron-absorbing screen around the fuel rod. The linear power of the fuel rod increased from 220 W/cm to 340 W/cm with a power ramp rate of 20 W/cm/min. The power of the fuel rod was monitored by both in-core thermal and nuclear measurement sensors in the test rig. This test provides experiences for further developing the power ramp test methods for PWR fuel rods at research reactor. (author)

  13. Study on B-10 consumption of PWR primary coolant during normal operation

    International Nuclear Information System (INIS)

    Liang, C.H.

    1994-01-01

    B-10 consumption under PWR primary coolant conditions has been analyzed. The result indicates its time-dependent change reacting with neutron in the normal operation. In this work, neutron energy assumed to be 4 eV; thermal neutron flux is in the range of 3 x 10 13 to 3 x 10 14 n/sec - cm 2 and the time of cycling of the primary coolant through the RCS is 8 sec. and its retention time in the core region is about 1 sec. Under this condition investigated, B-10 consumption is less than 5% at 3 x 10 13 n/sec - cm 2 thermal neutron flux, and closes to 27% at 3 x 10 14 n/sec - cm 2 by calculation at the 16th month of continuous operation. The effect of B-10 consumption on PWR primary water chemistry is also investigated. (author). 1 fig., 2 tabs., 4 refs

  14. Analysis of confinement effects for in-water seismic tests on PWR fuel assemblies

    International Nuclear Information System (INIS)

    Broc, Daniel; Queval, Jean-Claude; Rigaudeau, J.; Viallet, E.

    2001-01-01

    In the framework of a comprehensive program on the seismic behaviour of the PWR reactor cores, tests have been performed on a row of six PWR fuel assemblies, with two confinement configurations in water. Global fluid motion along the row is not allowed in the 'full confinement configuration', and is allowed in the 'lateral confinement configuration'. The seismic test results show that the impact forces at assembly grid levels are significantly smaller with the full confinement. This is due to damping, which is found to be larger in this configuration where the average fluid velocity inside the assembly (around the rods) is itself larger. We present analyses of these phenomena from theoretical and experimental standpoint. This involves both fluid models and structural models of the assembly row. (author)

  15. Computer codes for the study of the loss of coolant accident of PWR reactors

    International Nuclear Information System (INIS)

    Gomolinski, M.; Menessier, D.; Tellier, N.

    1975-01-01

    The CEA has undertaken a large programme to study the consequence on the core of the LOCA of a PWR. In the programme, simultaneously carried out experiments and the development of the calculations means are described. Several experiments such as OMEGA, ERSEC and PHEBUS tests, which provide data to check the computer codes are outlined briefly in the paper. For analysis of the LOCA of a PWR, a series of computer codes, which are at present in use or under development, are linked with each other. The codes are DANAIDES for blowdown, CERES for refill and reflood, THETA-1B and FLIRA for heat up calculation during the blow-down and the reflooding period respectively. FLIRA-PASTEL, a combination of FLIRA and PASTEL which calculate the stress and deformations of material using the finite element method, will be used in place of FLIRA. The basic models and flowcharts of the above codes are described in the paper

  16. German risk study of PWR's

    International Nuclear Information System (INIS)

    Kafka, P.

    1983-01-01

    In this paper, first the status of German Risk Study is presented briefly. Specific reference is made to the investigations in Phase B of the study and related programs. Significant elements involved in the risk assessment for NPPs, mainly in the field of system and structural reliability analyses are mentioned. In particular, important outcomes and limiting facts in the process of a Probabilistic Risk Assessment (PRA) to evaluate the safety standard and above all the influence of individual components or subsystems on core melt frequency are discussed. (orig.)

  17. Interface tracking simulations of bubbly flows in PWR relevant geometries

    Energy Technology Data Exchange (ETDEWEB)

    Fang, Jun, E-mail: jfang3@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Rasquin, Michel, E-mail: michel.rasquin@colorado.edu [Aerospace Engineering Department, University of Colorado, Boulder, CO 80309 (United States); Bolotnov, Igor A., E-mail: igor_bolotnov@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States)

    2017-02-15

    Highlights: • Simulations were performed for turbulent bubbly flows in PWR subchannel geometry. • Liquid turbulence is fully resolved by direct numerical simulation approach. • Bubble behavior is captured using level-set interface tracking method. • Time-averaged single- and two-phase turbulent flow statistical quantities are obtained. - Abstract: The advances in high performance computing (HPC) have allowed direct numerical simulation (DNS) approach coupled with interface tracking methods (ITM) to perform high fidelity simulations of turbulent bubbly flows in various complex geometries. In this work, we have chosen the geometry of the pressurized water reactor (PWR) core subchannel to perform a set of interface tracking simulations (ITS) with fully resolved liquid turbulence. The presented research utilizes a massively parallel finite-element based code, PHASTA, for the subchannel geometry simulations of bubbly flow turbulence. The main objective for this research is to demonstrate the ITS capabilities in gaining new insight into bubble/turbulence interactions and assisting the development of improved closure laws for multiphase computational fluid dynamics (M-CFD). Both single- and two-phase turbulent flows were studied within a single PWR subchannel. The analysis of numerical results includes the mean gas and liquid velocity profiles, void fraction distribution and turbulent kinetic energy profiles. Two sets of flow rates and bubble sizes were used in the simulations. The chosen flow rates corresponded to the Reynolds numbers of 29,079 and 80,775 based on channel hydraulic diameter (D{sub h}) and mean velocity. The finite element unstructured grids utilized for these simulations include 53.8 million and 1.11 billion elements, respectively. This has allowed to fully resolve all the turbulence scales and the deformable interfaces of individual bubbles. For the two-phase flow simulations, a 1% bubble volume fraction was used which resulted in 17 bubbles in

  18. TRANSPORT CHARACTERISTICS OF SELECTED PWR LOCA GENERATED DEBRIS

    International Nuclear Information System (INIS)

    MAJI, A. K.; MARSHALL, B.

    2000-01-01

    In the unlikely event of a Loss of Coolant Accident (LOCA) in a pressurized water reactor (PWR), break jet impingement would dislodge thermal insulation FR-om nearby piping, as well as other materials within the containment, such as paint chips, concrete dust, and fire barrier materials. Steam/water flows induced by the break and by the containment sprays would transport debris to the containment floor. Subsequently, debris would likely transport to and accumulate on the suction sump screens of the emergency core cooling system (ECCS) pumps, thereby potentially degrading ECCS performance and possibly even failing the ECCS. In 1998, the U. S. Nuclear Regulatory Commission (NRC) initiated a generic study (Generic Safety Issue-191) to evaluate the potential for the accumulation of LOCA related debris on the PWR sump screen and the consequent loss of ECCS pump net positive suction head (NPSH). Los Alamos National Laboratory (LANL), supporting the resolution of GSI-191, was tasked with developing a method for estimating debris transport in PWR containments to estimate the quantity of debris that would accumulate on the sump screen for use in plant specific evaluations. The analytical method proposed by LANL, to predict debris transport within the water that would accumulate on the containment floor, is to use computational fluid dynamics (CFD) combined with experimental debris transport data to predict debris transport and accumulation on the screen. CFD simulations of actual plant containment designs would provide flow data for a postulated accident in that plant, e.g., three-dimensional patterns of flow velocities and flow turbulence. Small-scale experiments would determine parameters defining the debris transport characteristics for each type of debris. The containment floor transport methodology will merge debris transport characteristics with CFD results to provide a reasonable and conservative estimate of debris transport within the containment floor pool and

  19. Invert Effective Thermal Conductivity Calculation

    International Nuclear Information System (INIS)

    M.J. Anderson; H.M. Wade; T.L. Mitchell

    2000-01-01

    The objective of this calculation is to evaluate the temperature-dependent effective thermal conductivities of a repository-emplaced invert steel set and surrounding ballast material. The scope of this calculation analyzes a ballast-material thermal conductivity range of 0.10 to 0.70 W/m · K, a transverse beam spacing range of 0.75 to 1.50 meters, and beam compositions of A 516 carbon steel and plain carbon steel. Results from this calculation are intended to support calculations that identify waste package and repository thermal characteristics for Site Recommendation (SR). This calculation was developed by Waste Package Department (WPD) under Office of Civilian Radioactive Waste Management (OCRWM) procedure AP-3.12Q, Revision 1, ICN 0, Calculations

  20. Inverted base pavements: construction and performance

    KAUST Repository

    Papadopoulos, Efthymios

    2017-05-11

    Inverted base pavements involve a well-compacted granular aggregate base built between a thin asphalt concrete layer and a cement-treated base. Inverted base pavements can be constructed using conventional equipment and procedures but require proper quality control. This study reviews the extensive South African experience and case histories in the USA. Accumulating evidence suggests that inverted base pavements are a viable alternative and can outperform conventional pavements at a lower cost. Inverted base pavements rely on the complementary interaction between layers. The cement-treated base provides a stiff foundation for efficient compaction and constrains the deformation of the stress-sensitive granular aggregate base. The thin asphalt surface layer deforms as a membrane and develops low tensile stress. Additional large-scale field tests should be conducted to assess the performance of inverted base pavement designs in a wide range of conditions relevant to the USA.

  1. On Invertible Sampling and Adaptive Security

    DEFF Research Database (Denmark)

    Ishai, Yuval; Kumarasubramanian, Abishek; Orlandi, Claudio

    2011-01-01

    functionalities was left open. We provide the first convincing evidence that the answer to this question is negative, namely that some (randomized) functionalities cannot be realized with adaptive security. We obtain this result by studying the following related invertible sampling problem: given an efficient...... sampling algorithm A, obtain another sampling algorithm B such that the output of B is computationally indistinguishable from the output of A, but B can be efficiently inverted (even if A cannot). This invertible sampling problem is independently motivated by other cryptographic applications. We show......, under strong but well studied assumptions, that there exist efficient sampling algorithms A for which invertible sampling as above is impossible. At the same time, we show that a general feasibility result for adaptively secure MPC implies that invertible sampling is possible for every A, thereby...

  2. Analysis of core and core barrel heat-up under conditions simulating severe reactor accidents

    International Nuclear Information System (INIS)

    Chellaiah, S.; Viskanta, R.; Ranganathan, P.; Anand, N.K.

    1987-01-01

    This paper reports on the development of a model for estimating the temperature distributions in the reactor core, core barrel, thermal shield and reactor pressure vessel of a PWR during an undercooling transient. A number of numerical calculations simulating the core uncovering of the TMI-2 reactor and the subsequent heat-up of the core have been performed. The results of the calculations show that the exothermic heat release due to Zircaloy oxidation contributes to the sharp heat-up of the core. However, the core barrel temperature rise which is driven by the temperature increase of the edge of the core (e.g., the core baffle) is very modest. The maximum temperature of the core barrel never exceeded 610 K (at a system pressure of 68 bar) after a 75 minute simulation following the start of core uncovering

  3. Flow visualization study of inverted annular flow of post dryout heat transfer region

    International Nuclear Information System (INIS)

    Ishii, M.; De Jarlais, G.

    1987-01-01

    The inverted annular flow is important in the area of LWR accident analysis in terms of the maximum cladding temperature and effectiveness of the emergency core cooling. However, the inverted annular flow thermal-hydraulics is not well understood due to its special heat transfer condition of film boiling. In view of this, the inverted flow is studied in detail experimentally. A new experimental apparatus has been constructed in which film boiling heat transfer can be established in a transparent test section. Data on liquid core stability, core break-up mechanism, and dispersed-core liquid slug and droplet sizes are obtained using F 113 as a test fluid. Both high speed movies and flash photographs are used. The inlet section consists of specially designed coaxial nozzles for gas and liquid such that the ideal inverted annular flow can be generated. The roll wave formation, droplet entrainment from wave crests, agitated sections with large interfacial areas, classical sinuous jet instability, jet break-up into multiple liquid ligaments and drop formation from liquid ligaments have been observed in detail. (orig.)

  4. Technical Support to an Operating PWR vis-a-vis Safety Analysis

    International Nuclear Information System (INIS)

    Gul, Subhan; Khan, M.; Chughtai, M. Kamran

    2011-01-01

    Currently a PWR of 300 MWe capacity CHASNUPP-I is in operation since the year 2000. Technical support being provided includes in-core fuel management and corresponding safety analysis for the reshuffled core for the next cycle. Currently calculation and analysis were performed for Cycle 6 to achieve the safe and economical loading pattern. The technique used is designated as out in mode (modified). In this technique, most of the fresh fuel assemblies are not directly located at the periphery of the core, but near the boundary. This technique has the advantage that without using burnable absorber we can design a low leakage core with extended cycle and maximum batch averaged burnup. (author)

  5. 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

  6. 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

  7. 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

  8. Radiation detectors for the control of PWR nuclear boilers

    International Nuclear Information System (INIS)

    Duchene, J.

    1977-01-01

    The neutronic control in French PWR is effected by: 2 channels of measurement of intermediate power using γ'-compensated boron-coated ionization chambers 4 channels of measurement of high power with 'long' boron chambers also used in axial off-set measurement. A movable in-core measuring system is used for the fuel management and the power distribution monitoring. The instrumentation of start-up and intermediate power is conventional; the chambers of the axial off-set measurement and the in-core system are special for this type of power plant, they are discussed in details. The essential properties of the various types of detector, their major advantages or drawbacks, their comparative adaptation to the functions to be performed in the plant are summarized in a table. The 'long chambers' (on use in Fessenheim I and II, and soon in Bugey II) are boron coated current ionization chambers, without γ compensation, intended for power measurement. In-core measurements first involved activation methods - movable wires giving flux profiles, -or activable nuts (the Aeroball System at Trino Vercellese, Chooz...). In on-line neutron detectors, used at fixed positions, the electric signal is generated from: ionization the gas filling fission ionization chambers and γ ionization chambers; direct collection of the charged particles emitted from the convertor element in self-powered neutron detectors (rhodium, silver or vanadium) or self-powered γ detectors (cobalt); or thermoelectric effect in neutron and γ thermometers. The in-core measurement unit developped by Framatome is a movable miniaturized fission chamber system (at Tihange), every French exported power plant being now equipped with it [fr

  9. Analyses of PWR boron dilution consequences with the Arrotta code

    International Nuclear Information System (INIS)

    Johanson, E.; Cheng, H.W.; Sehgal, B.R.

    1998-03-01

    During the past few years, major attention has been paid to analyzing the issue of reactivity initiated accidents (RIAs), of which the boron dilution event is of very special interest to the countries having pressurized water reactors (PWRs) in their nuclear power delivery systems. The scenario considered is that if an inadvertent accumulation of boron free water in one loop during reactor startup operations of a PWR and the inadvertent startup of the reactor coolant pump (RCP) in the loop. This could then lead to a rapid boron dilution in the core, which can in turn give rise to a power excursion. This report is devoted to studying the potential physical and thermal hydraulic consequences of a slug of diluted coolant entering the core after one RCP start under a couple of postulated cases. The severity of the consequences of such a scenario is primarily determined by the amount of positive reactivity insertion, and they are also related to the reactivity insertion rate. Therefore, in the report, detailed calculations and analyses have been carried out from case to case by using the well-known space-time kinetics code, ARROTTA. As a result, the spatial distribution for nodal power, fuel enthalpy, fuel temperature and clad outside temperature as well as the change in core reactivity, total core power and peak fuel temperature can be provided. In general, the maximum fuel enthalpy, peak fuel temperature, and clad outside temperature, for all the cases considered in the report, do not exceed their respective routine safety limitations because of the strong Doppler effect and moderator temperature feedback, except if the safety limitations on fuel enthalpy addition for high burnup fuel are drastically reduced

  10. Seismic behaviour of PWR fuel assemblies model and its validation

    International Nuclear Information System (INIS)

    Queval, J.C.; Gantenbein, F.; Brochard, D.; Benjedidia, A.

    1991-01-01

    The validity of the models simulating the seismic behaviour of PWR cores can only be exactly demonstrated by seismic testing on groups of fuel assemblies. Shake table seismic tests of rows of assembly mock-ups, conducted by the CEA in conjunction with FRAMATOME, are presented in reference /1/. This paper addresses the initial comparisons between model and test results for a row of five assemblies in air. Two models are used: a model with a single beam per assembly, used regularly in accident analyses, and described in reference /2/, and a more refined 2-beam per assembly model, geared mainly towards interpretation of test results. The 2-beam model is discussed first, together with parametric studies used to characterize it, and the study of the assembly row for a period limited to 2 seconds and for different excitation levels. For the 1-beam model assembly used in applications, the row is studied over the total test time, i.e twenty seconds, which covers the average duration of the core seismic behaviour studies, and for a peak exciting acceleration value at 0.4 g, which corresponds to the SSE level of the reference spectrum

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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.)

  16. 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)

  17. 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

  18. 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.)

  19. 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)

  20. 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