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Sample records for advanced pwr core

  1. Conceptual study of advanced PWR core design

    International Nuclear Information System (INIS)

    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

  2. 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.

  3. Conceptual study of advanced PWR core design. Development of advanced PWR core neutronics analysis system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Hyo; Kim, Seung Cho; Kim, Taek Kyum; Cho, Jin Young; Lee, Hyun Cheol; Lee, Jung Hun; Jung, Gu Young [Seoul National University, Seoul (Korea, Republic of)

    1995-08-01

    The neutronics design system of the advanced PWR consists of (i) hexagonal cell and fuel assembly code for generation of homogenized few-group cross sections and (ii) global core neutronics analysis code for computations of steady-state pin-wise or assembly-wise core power distribution, core reactivity with fuel burnup, control rod worth and reactivity coefficients, transient core power, etc.. The major research target of the first year is to establish the numerical method and solution of multi-group diffusion equations for neutronics code development. Specifically, the following studies are planned; (i) Formulation of various numerical methods such as finite element method(FEM), analytical nodal method(ANM), analytic function expansion nodal(AFEN) method, polynomial expansion nodal(PEN) method that can be applicable for the hexagonal core geometry. (ii) Comparative evaluation of the numerical effectiveness of these methods based on numerical solutions to various hexagonal core neutronics benchmark problems. Results are follows: (i) Formulation of numerical solutions to multi-group diffusion equations based on numerical methods. (ii) Numerical computations by above methods for the hexagonal neutronics benchmark problems such as -VVER-1000 Problem Without Reflector -VVER-440 Problem I With Reflector -Modified IAEA PWR Problem Without Reflector -Modified IAEA PWR Problem With Reflector -ANL Large Heavy Water Reactor Problem -Small HTGR Problem -VVER-440 Problem II With Reactor (iii) Comparative evaluation on the numerical effectiveness of various numerical methods. (iv) Development of HEXFEM code, a multi-dimensional hexagonal core neutronics analysis code based on FEM. In the target year of this research, the spatial neutronics analysis code for hexagonal core geometry(called NEMSNAP-H temporarily) will be completed. Combination of NEMSNAP-H with hexagonal cell and assembly code will then equip us with hexagonal core neutronics design system. (Abstract Truncated)

  4. Advanced methods for the study of PWR cores

    International Nuclear Information System (INIS)

    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.)

  5. Neutronic Analysis of Advanced SFR Burner Cores using Deep-Burn PWR Spent Fuel TRU Feed

    International Nuclear Information System (INIS)

    In this work, an advanced sodium-cooled fast TRU (Transuranics) burner core using deep-burn TRU feed composition discharged from small LWR cores was neutronically analyzed to show the effects of deeply burned TRU feed composition on the performances of sodium-cooled fast burner core. We consider a nuclear park that is comprised of the commercial PWRs, small PWRs of 100MWe for TRU deep burning using FCM (Fully Ceramic Micro-encapsulated) fuels and advanced sodium-cooled fast burners for their synergistic combination for effective TRU burning. In the small PWR core having long cycle length of 4.0 EFPYs, deep burning of TRU up to 35% is achieved with FCM fuel pins whose TRISO particle fuels contain TRUs in their central kernel. In this paper, we analyzed the performances of the advanced SFR burner cores using TRU feeds discharged from the small long cycle PWR deep-burn cores. Also, we analyzed the effect of cooling time for the TRU feeds on the SFR burner core. The results showed that the TRU feed composition from FCM fuel pins of the small long cycle PWR core can be effectively used into the advanced SFR burner core by significantly reducing the burnup reactivity swing which reduces smaller number of control rod assemblies to satisfy all the conditions for the self controllability than the TRU feed composition discharged from the typical PWR cores

  6. Overview of CEA studies on advanced Plutonium fuelled PWR core concepts

    International Nuclear Information System (INIS)

    After a brief summary of the French experience of Plutonium recycle in PWRs and a review of the rationale for this strategy, a first part of the paper presents the plans for generalizing the recycling of 30% MOX in present generating facilities and the next generation reactors (EPR). A second part of the paper is dedicated to an overview of CEA studies on advanced 100% MOX PWR core concepts compatible with the EPR design and likely to improve the performances of the reference core design in terms of Plutonium consumption, and minor actinides production. The respective merits of these advanced 100% MOX PWR core concepts are briefly summarized, as well as their potential benefits to accommodate a wide variety of future options for the fuel cycle. (author) 3 figs., 6 tabs., 11 refs

  7. PWR degraded core analysis

    International Nuclear Information System (INIS)

    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.)

  8. 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.)

  9. PWR core monitoring system and benchmarking

    International Nuclear Information System (INIS)

    The PWR Power Shape Monitoring System (PSMS) provides site engineers with new capabilities for monitoring and predicting core power distributions. These capabilities can lead to increased plant output as a result of greater operating margins, better load maneuvering, earlier detection of anomalies, and improved fuel reliability. The heart of the PSMS consists of nodal code (NODEP-2/THERM-P) that computes the 3-D core power distribution. This code is coupled to a simplified nodal version of the COBRA-IIIC/MIT-2 thermal-hydraulic model to determine the DNBR. These calculations can be completed in about 30 seconds on a PRIME-750 mini computer. Activation of the calculations and review of the results is through user-friendly interactive software that can be tailored to the requirements and capabilities of the different categories of users through table-driven menus. The PSMS provides unique advances over core power monitoring systems based purely on measurements. The PSMS approach permits the three-dimensional core simulation model to be routinely corrected with in-core/ex-core measurements while simultaneously identifying consistent instrument errors

  10. Full MOX core design for PWR

    International Nuclear Information System (INIS)

    Full MOX core design for APWR was analyzed in nuclear design, fuel integrity analysis, thermal hydraulic design and safety analysis et. al. Feasibility of Full MOX core was confirmed from these analyses without any large modifications. Full MOX PWR core has very good characteristics in which single Pu content in an assembly, burnable poison free, higher burnup and longer cycle operation are feasible. (author)

  11. Thermohydraulic and constructional boundary conditions of an advanced PWR reactor

    International Nuclear Information System (INIS)

    The advantages and special features of an advanced PWR reactor (FDWR) have been systematically investigated for several years by the Department of Space Flight and Reactor Technology of the University of Brunswick (LRR-TUBS). The FDWR will have a homogeneous core, i.e. the fuel elements will consist of fuel rods of the same size and enrichment. (orig./GL)

  12. PWR Core 2 Project accident analysis

    International Nuclear Information System (INIS)

    The various operations required for receipt, handling, defueling and storage of spent Shippingport PWR Core 2 fuel assemblies have been evaluated to determine the potential accidents and their consequences. These operations will introduce approximately 16,500 kilograms of depleted natural uranium (as UO2), 139 kilograms of plutonium, 2.8 megacuries of mixed fission products, and 14 kilograms of Zircaloy-4 (cladding and hardware) into the 221-T Canyon Building. Event sequences for potential accidents that were considered included (1) leaking fuel assemblies, (2) fire and explosion, (3) loss of coolant or cooling capability, (4) dropped and/or damaged fuel assemblies, and extrinsic occurrences such as loss of services, missile impact, and natural occurrences (e.g., earthquake, tornado). Accident frequencies were determined by formal analysis to be very low. Accident consequences are greatly mitigated by the safety and containment features designed into the fuel modules and shipping cask, the long cooling time since reactor discharge, and the redundant safety features designed into the facilities, equipment, and operating procedures for the PWR Core 2 Project. Possible hazards associated with the handling of these fuels have been considered and adequate safeguards and storage constraints identified. The operations of M-160 cask unloading and module storage will not involve identifiable risks as great or significantly greater than those for comparable licensed nuclear facilities, nor will hazards or risks be significantly different from comparable past 221-T Plant programs. Therefore, it is concluded that the operations required for receipt, handling, and defueling of the M-160 cask and for the storage and surveillance of the PWR Core 2 fuel assemblies at the 221-T Canyon Building can be performed without undue risk to the safety of the involved personnel, the public, the environment or the facility

  13. Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR/CANDU synergism

    Institute of Scientific and Technical Information of China (English)

    HUO Xiao-Dong; XIE Zhong-Sheng

    2004-01-01

    High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexibility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nuclear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (~22.5%), increase the energy output (~41%), decrease the quantity of spent fuels to be disposed (~2/3) and lower the cost of nuclear power. Because of the inherent flexibility of nuclear fuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modification of the reactor core structure and operation mode. It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.

  14. Evaluation of tight-pitch PWR cores

    International Nuclear Information System (INIS)

    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 - 235U/UO2 : Pu/ThO2 : 233U/ThO2 - and the conventional recycle-mode uranium system - 235U/UO2 : Pu/UO2. 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

  15. VERA Core Simulator Methodology for PWR Cycle Depletion

    Energy Technology Data Exchange (ETDEWEB)

    Kochunas, Brendan [University of Michigan; Collins, Benjamin S [ORNL; Jabaay, Daniel [University of Michigan; Kim, Kang Seog [ORNL; Graham, Aaron [University of Michigan; Stimpson, Shane [University of Michigan; Wieselquist, William A [ORNL; Clarno, Kevin T [ORNL; Palmtag, Scott [Core Physics, Inc.; Downar, Thomas [University of Michigan; Gehin, Jess C [ORNL

    2015-01-01

    This paper describes the methodology developed and implemented in MPACT for performing high-fidelity pressurized water reactor (PWR) multi-cycle core physics calculations. MPACT is being developed primarily for application within the Consortium for the Advanced Simulation of Light Water Reactors (CASL) as one of the main components of the VERA Core Simulator, the others being COBRA-TF and ORIGEN. The methods summarized in this paper include a methodology for performing resonance self-shielding and computing macroscopic cross sections, 2-D/1-D transport, nuclide depletion, thermal-hydraulic feedback, and other supporting methods. These methods represent a minimal set needed to simulate high-fidelity models of a realistic nuclear reactor. Results demonstrating this are presented from the simulation of a realistic model of the first cycle of Watts Bar Unit 1. The simulation, which approximates the cycle operation, is observed to be within 50 ppm boron (ppmB) reactivity for all simulated points in the cycle and approximately 15 ppmB for a consistent statepoint. The verification and validation of the PWR cycle depletion capability in MPACT is the focus of two companion papers.

  16. Advanced PWR fuel design concepts

    International Nuclear Information System (INIS)

    For nearly 15 years, Combustion Engineering has provided pressurized water reactor fuel with the features most suppliers are now introducing in their advanced fuel designs. Zircaloy grids, removable upper end fittings, large fission gas plenum, high burnup, integral burnable poisons and sophisticated analytical methods are all features of C-E standard fuel which have been well proven by reactor performance. C-E's next generation fuel for pressurized water reactors features 24-month operating cycles, optimal lattice burnable poisons, increased resistance to common industry fuel rod failure mechanisms, and hardware and methodology for operating margin improvements. Application of these various improvements offer continued improvement in fuel cycle economics, plant operation and maintenance. (author)

  17. 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.

  18. An evaluation of tight - pitch PWR cores

    International Nuclear Information System (INIS)

    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.)

  19. Zebra: An advanced PWR lattice code

    Energy Technology Data Exchange (ETDEWEB)

    Cao, L.; Wu, H.; Zheng, Y. [School of Nuclear Science and Technology, Xi' an Jiaotong Univ., No. 28, Xianning West Road, Xi' an, ShannXi, 710049 (China)

    2012-07-01

    This paper presents an overview of an advanced PWR lattice code ZEBRA developed at NECP laboratory in Xi'an Jiaotong Univ.. The multi-group cross-section library is generated from the ENDF/B-VII library by NJOY and the 361-group SHEM structure is employed. The resonance calculation module is developed based on sub-group method. The transport solver is Auto-MOC code, which is a self-developed code based on the Method of Characteristic and the customization of AutoCAD software. The whole code is well organized in a modular software structure. Some numerical results during the validation of the code demonstrate that this code has a good precision and a high efficiency. (authors)

  20. Zebra: An advanced PWR lattice code

    International Nuclear Information System (INIS)

    This paper presents an overview of an advanced PWR lattice code ZEBRA developed at NECP laboratory in Xi'an Jiaotong Univ.. The multi-group cross-section library is generated from the ENDF/B-VII library by NJOY and the 361-group SHEM structure is employed. The resonance calculation module is developed based on sub-group method. The transport solver is Auto-MOC code, which is a self-developed code based on the Method of Characteristic and the customization of AutoCAD software. The whole code is well organized in a modular software structure. Some numerical results during the validation of the code demonstrate that this code has a good precision and a high efficiency. (authors)

  1. Design of a PWR emergency core cooling simulator loop

    International Nuclear Information System (INIS)

    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)

  2. Westinghouse advanced passive 600 MWe PWR design

    International Nuclear Information System (INIS)

    Although there has been a sharp downturn in the ordering of commercial nuclear power plants throughout the world, it is nonetheless anticipated that this form of energy will remain vital to the economy of many nations in a long term. One of the important new development activities is that of small plants incorporating passive safety features. The small plants have the merits in terms of low total capital requirement and potentially short lead time. The Electric Power Research Institute sponsored the development of an advanced LWR plant in a nine month Westinghouse program, which terminated in March, 1986. Further development at Westinghouse is now in progress on this design called AP 600 under the sponsorship of the U.S. Department of Energy. On the basis of the proven 600 MWe PWR plant design, the specific design improvement for increased safety and operational margin, reduced plant capital and operating cost, simplified plant systems and components, and increased certainty of meeting construction schedule and cost is pursued. The Westinghouse two-loop plants are very competitive, and the operating performance is outstanding by the comparison of plant capacity factor. The operation and maintenance costs are low. The specific design and the features of modification and improvement are discussed. (Kako, I)

  3. Conceptual study on advanced PWR system

    International Nuclear Information System (INIS)

    In this study, the adoptable essential technologies and reference design concept of the advanced reactor were developed and related basic experiments were performed. 1) Once-through Helical Steam Generator: a performance analysis computer code for heli-coiled steam generator was developed for thermal sizing of steam generator and determination of thermal-hydraulic parameters. 2) Self-pressurizing pressurizer : a performance analysis computer code for cold pressurizer was developed. 3) Control rod drive mechanism for fine control : type and function were surveyed. 4) CHF in passive PWR condition : development of the prediction model bundle CHF by introducing the correction factor from the data base. 5) Passive cooling concepts for concrete containment systems: development of the PCCS heat transfer coefficient. 6) Steam injector concepts: analysis and experiment were conducted. 7) Fluidic diode concepts : analysis and experiment were conducted. 8) Wet thermal insulator : tests for thin steel layers and assessment of materials. 9) Passive residual heat removal system : a performance analysis computer code for PRHRS was developed and the conformance to EPRI requirement was checked. (author). 18 refs., 55 tabs., 137 figs

  4. Endurance test for non-instrumented capsule of advanced PWR fuel pellet (test procedure)

    International Nuclear Information System (INIS)

    This test procedure details the test loop, test method, and test procedure for pressure drop, vibration and endurance test of Non-instrumented Capsule of Advanced PWR Fuel Pellet. From the pressure drop test, the hydraulic design requirements of the capsule are verified. HANARO limit condition is checked and the compatibility with HANARO core is verified. From flow induced vibration test vibration frequency, vibration displacement are investigated. The wear of Non-instrumented Capsule of Advanced PWR Fuel Pellet is investigated through endurance test, and these data are used to evaluate the expected wear of during maximum resident time of Non-instrumented Capsule

  5. DNB experiments for high-conversion PWR core design

    International Nuclear Information System (INIS)

    It is very important to clarify the departure from nucleate boiling (DNB) performance of core fuel assemblies for the high conversion pressurized water reactors (PWR). To investigate this, DNB experiments were performed in tight lattice rod bundles, using the model fluid Freon 12 and water under the actual operating conditions. In addition, DNB heat flux measurements in an annular-flow channel were carried out for the design of the fertile rods, which are installed in thimble tubes. (orig.)

  6. DNB experiments for high conversion PWR core design

    International Nuclear Information System (INIS)

    It is very important to clarify the departure from nucleate boiling (DNB) performance of core fuel assemblies for the high conversion PWR design. To investigate this, DNB experiments were performed in tight lattice rod bundles, using the model fluid freon-12 and the actual water. And also DNB heat flux mesurements in an annular flow channel were carried out for design of fertile rods which are installed in thimble tubes. (orig.)

  7. Three dimensions transport calculations for PWR core

    International Nuclear Information System (INIS)

    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)

  8. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    International Nuclear Information System (INIS)

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10−6 on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure

  9. Evolutionary developments of advanced PWR nuclear fuels and cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyu-Tae, E-mail: ktkim@dongguk.ac.kr

    2013-10-15

    Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10{sup −6} on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure.

  10. Axial simulation of PWR core and study of actuators

    International Nuclear Information System (INIS)

    Development of an operation code allowing to simulate the behaviour of a PWR type reactor core. Load following is controled by bore and control rods, taking into account the temperature counter-reactions. The fine behaviour of the fuel element during transients is not simulated, on the other hand the central part of the reactor is completely simulated. The regulation equation are easily modifiable and thus it is possible to test in open loop any modification brought about to this regulation. Description of simulation tests on CAS-2B reactor: core control, static tests, dynamic tests

  11. Supernova - the multi-dimensional core model of the Westinghouse on-line PWR core monitor, BEACON

    International Nuclear Information System (INIS)

    BEACON is an advanced multi-dimensional on-line PWR core monitoring system developed at Westinghouse. It resides on a workstation, and performs a large variety of very fast and accurate three-dimensional core analyses. The high speed performance of the BEACON system is made possible through the use of a superfast multi-dimensional nodal code, SUPERNOVA (SPNOVA), which is recently developed at Westinghouse. Compared to conventional nodal codes, SPNOVA is two orders of magnitude faster and yields predictions of comparable accuracy. This article describes the methodology of SPNOVA and presents examples of qualification data for both SPNOVA and BEACON

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

    International Nuclear Information System (INIS)

    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)

  13. Development and preliminary verification of the PWR on-line core monitoring software system. SOPHORA

    International Nuclear Information System (INIS)

    This paper presents an introduction to the development and preliminary verification of a new on-line core monitoring software system (CMSS), named SOPHORA, for fixed in-core detector (FID) system of PWR. Developed at China General Nuclear Power Corporation (CGN), SOPHORA integrates CGN’s advanced PWR core simulator COCO and thermal-hydraulic sub-channel code LINDEN to manage the real-time core calculation and analysis. Currents measured by the FID are re-evaluated and used as bases to reconstruct the 3-D core power distribution. The key parameters such as peak local power margin and minimum DNBR margin are obtained by comparing with operation limits. Pseudo FID signals generated by data from movable in-core detector (MID) are used to verify the SOPHORA system. Comparison between predicted power peak and the responding MID in-core flux map results shows that the SOPHORA results are reasonable and satisfying. Further verification and validation of SOPHORA is undergoing and will be reported later. (author)

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

    International Nuclear Information System (INIS)

    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

  15. The core-wise green's function method and its application to in-core fuel management and reload optimization for PWR

    International Nuclear Information System (INIS)

    A fuel management code CGFMAC and a optimization code OPCGFM for PWR are encoded with the Core-Wise Green's Function Method (CGFM). They are qualified by Qinshan Nuclear Power Plant's problems. The numerical results demonstrate that the CGFM method is about 10 times faster than the advanced nodal Green's function method such as NGFM, and about 20 times faster than the coarse mesh difference method. All results calculated with these different methods agree well with each other. It is demonstrated that the CGFMAC and OPCGFM can be used in the core design for PWR

  16. Passive containment cooling for an advanced PWR

    International Nuclear Information System (INIS)

    The AP600 is a 600-MW(electric) pressurized water reactor that is currently being developed by Westinghouse and its subcontractors. The AP600 program is being sponsored by the US Department of Energy (DOE) in conjunction with DOE and Electric Power Research Institute advanced light water reactor programs. The AP600 employs safety features that, when actuated, use natural phenomena and stored energy (gravity, natural circulation, compressed gas) to accomplish all required safety functions. This safety approach results in both improved safety and a significant simplification in the overall plant design since no safety-grade ac power or support systems are required. Also, significant reductions in plant complexity, capital cost, and construction schedule can be achieved. One of the key safety systems in the AP600 passive safety approach is the passive containment cooling system (PCCS). The PCCS provides the safety-grade ultimate heat sink for the removal of reactor-sensible heat and core decay heat following any design-basis event. Analytical models of the PCCS have been developed and transient and accident evaluations have been performed to demonstrate the heat removal capability to the PCCS. These analyses indicate that AP600 postaccident containment response is similar to that achieved with active containment heat removal systems. Also in conjunction with analyses, a test program is under way to demonstrate and verify the heat removal capability of the PCCS design concept

  17. Simulation of fission products behavior in severe accidents for advanced passive PWR

    International Nuclear Information System (INIS)

    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

  18. A highly heterogeneous 3D PWR core benchmark: deterministic and Monte Carlo method comparison

    International Nuclear Information System (INIS)

    Physical analyses of the LWR potential performances with regards to the fuel utilization require an important part of the work dedicated to the validation of the deterministic models used for theses analyses. Advances in both codes and computer technology give the opportunity to perform the validation of these models on complex 3D core configurations closed to the physical situations encountered (both steady-state and transient configurations). In this paper, we used the Monte Carlo Transport code TRIPOLI-4 to describe a whole 3D large-scale and highly-heterogeneous LWR core. The aim of this study is to validate the deterministic CRONOS2 code to Monte Carlo code TRIPOLI-4 in a relevant PWR core configuration. As a consequence, a 3D pin by pin model with a consistent number of volumes (4.3 millions) and media (around 23.000) is established to precisely characterize the core at equilibrium cycle, namely using a refined burn-up and moderator density maps. The configuration selected for this analysis is a very heterogeneous PWR high conversion core with fissile (MOX fuel) and fertile zones (depleted uranium). Furthermore, a tight pitch lattice is selected (to increase conversion of 238U in 239Pu) that leads to harder neutron spectrum compared to standard PWR assembly. This benchmark shows 2 main points. First, independent replicas are an appropriate method to achieve a fare variance estimation when dominance ratio is near 1. Secondly, the diffusion operator with 2 energy groups gives satisfactory results compared to TRIPOLI-4 even with a highly heterogeneous neutron flux map and an harder spectrum

  19. The development of 1530 MW steam turbine for Advanced PWR

    International Nuclear Information System (INIS)

    MITSUBISHI has been manufacturing 27 nuclear steam turbines and total output is over 20,000 MW since the 1970 first delivery of nuclear steam turbine. Based on these our successful experiences, MITSUBISHI is making a continuous effort to develop the most modern steam turbine with the lager capacity, higher efficiency and higher reliability. And now, the first Advanced PWR is being planned to be built at Tsuruga No.3 and No.4 by Japan Atomic Power Co. as the largest plant with an electric power of about 1530 MW. To apply this Advanced PWR plants, we are going forward planing and developing the largest capacity nuclear steam turbine. This paper shows the key technologies of target capacity nuclear steam turbine such as 54 inches low pressure last blade, and the advanced technologies to realize high performance and high reliability steam turbine. (author)

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

    International Nuclear Information System (INIS)

    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)

  1. Modelling of whole-core release of fission products in PWR core melt accidents: Chapter 13

    International Nuclear Information System (INIS)

    The computer code FISREL combines the thermal history of a reactor core with experimentally-based release rate constants to calculate whole-core release histories of fission products in PWR core melt accidents. Predictions of the code for releases of volatile fission products during large-break, small-break and transient initiated sequences are presented, and the sensitivities of results to input data examined. A preliminary assessment of the limitations imposed by mass transport on release of vaporized materials in high pressure sequences is given, and the implications of the results for primary system transport are discussed

  2. Assessment of spectral history influence on PWR and WWER core

    International Nuclear Information System (INIS)

    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. Neglecting this effect leads to an additional component of error in neutron-physical characteristics. Two solution approaches to this problem implemented in the reactor dynamic code DYN3D are described and compared in this paper: a cross section correction method based on 239Pu concentration and separate cross sections treatment for each axial layer of reactor core. Steady-state and burnup characteristics of a PWR and a WWER-1000 cores, calculated by DYN3D with and without cross section corrections, are compared. An impact of the correction on transient calculations is studied for a control rod ejection example. Studies have shown a significant influence of spectral history on axial power and burnup distributions as well as on calculated cycle length. Two different correction methods have shown similar major effects. (orig.)

  3. 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.

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

    International Nuclear Information System (INIS)

    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)

  5. Development of advanced PWR system analysis technology

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Y. D.; Kim, S. O.; Jung, B. D.; Kim, Y. I.; Chang, M. H.; Lee, Y. J.; Yun, J. H.

    1997-12-31

    The scope of this project is to establish the basic analysis technologies for the advanced designed with the passive and inherent safety concepts. The scope is extended to the application of these technologies to the performance and safety analysis of the passive reactor. Since the different design concepts are applied depending on the reactor power, the study is conducted for the small and medium sized integral reactor as well as the large scale passive reactors by focusing on the analysis technology development for the passive components. The design concepts which can be applied for the safety enhancement of the domestic advanced reactor are developed through evaluating the technical information of the overseas advanced reactor concepts.

  6. Experimental research progress on passive safety systems of Chinese advanced PWR

    International Nuclear Information System (INIS)

    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

  7. Coupled neutronics and thermal-hydraulic solution of a full-core PWR using VERA-CS

    International Nuclear Information System (INIS)

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing a core simulator called VERA-CS to model operating pressurized water reactors (PWRs) with high resolution. This paper describes how the development of VERA-CS is being driven by a set of progression benchmark problems that specify the delivery of useful capability in discrete steps. As part of this development, this paper will describe the current capability of VERA-CS to perform a multiphysics simulation of an operating PWR at Hot Full Power (HFP) conditions using a set of existing computer codes coupled together in a novel method. Results for several single-assembly cases are shown that demonstrate coupling for different boron concentrations and power levels. Finally, high-resolution results are shown for a full-core PWR reactor modeled in quarter-symmetry. (author)

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

    International Nuclear Information System (INIS)

    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

  9. Core power distribution methodology in the BEACON PWR [pressurized water reactor] core monitoring system

    International Nuclear Information System (INIS)

    Westinghouse has developed an advanced operational core support package called BEACON which uses a fully analytical methodology for on-line prediction of 3-D [three-dimensional] power distributions. The system provides core monitoring, core measurement reduction, core analysis and follow, and core predictions. The heart of the system is a very fast and accurate three dimensional nodal code which is used for core simulation and predictions. The system uses a new methodology with the existing core instrumentation to infer the current measured power distribution. This methodology has been qualified and yields excellent results

  10. Advanced Core Monitoring Framework: An overview description

    International Nuclear Information System (INIS)

    One of the most significant developments in nuclear power plant operations in recent years is the application of digital computers to monitor and manage power plant process. The introduction of this technology, moreover is not without its problems. At present each of these advanced core monitoring systems as GE's MONICORE, EXXON's POWERPLEX, EPRI's PSMS, etc., works only by itself in an operating configuration which makes it difficult to compare, benchmark or replace with alternative core monitoring packages. The Advanced Core Monitoring Framework (ACMF) was conceived to provide one standard software framework in a number of different virtual-memory mini-computers within which modules from any of the core monitoring systems (both BWR and PWR) could be installed. The primary theme of ACMF is to build a framework that allows software plug-in compatibility for a variety of core monitoring functional packages by carefully controlling (standardizing) module interfaces to a well-defined database and requiring a common man-machine interface to be installed

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

    Energy Technology Data Exchange (ETDEWEB)

    Papukchiev, Angel [GRS mbH Forschungsinstitute, Garching (Germany); Schaefer, Anselm [ISaR GmbH, Garching (Germany)

    2008-07-01

    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

  12. Development of in-core fuel management scoping tools for PWR

    International Nuclear Information System (INIS)

    This paper concerns with developing a simplified in-core fuel management scoping tool for PWR. For this purpose the point reactivity model is put into a fuel cycling decision code, FCYPRM. Modified Borresen's coarse-mesh diffusion theory and nodal expansion method are utilized to form a spatial neutron analysis code, CMSNAP. Numerical experiments are performed to determine a set of empirical shuffling rules for working out an automated fuel loading pattern search code, ALPS. The numerical examples are presented for verifying effectiveness and applicability of individual codes. By structuring and applying three codes for reload core design problem of a PWR, it is demonstrated that these codes provide an effective in-core fuel management scoping tool for PWR. (Author)

  13. Intelligent main control room for advanced PWR plants

    International Nuclear Information System (INIS)

    The design targets of the main control room of nuclear power plants are as follows. (1) To make a good working environment where operators can operate easily. (2) To reduce the work load and operators error. To this end, MHI has been improving main control room design for advanced PWR plants. The new intelligent main control room consists of a soft operation console and a large display panel. According to our evaluation, the work load and human error of the new main control room are reduced by about 35% compared with the latest plants. This new design will be used to plan new plants and will have the additional feature of saving costs by standardizing plant design. (author)

  14. Advanced PWR technology development -Development of advanced PWR system analysis technology-

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Moon Heui; Hwang, Yung Dong; Kim, Sung Oh; Yoon, Joo Hyun; Jung, Bub Dong; Choi, Chul Jin; Lee, Yung Jin; Song, Jin Hoh [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    The primary scope of this study is to establish the analysis technology for the advanced reactor designed on the basis of the passive and inherent safety concepts. This study is extended to the application of these technology to the safety analysis of the passive reactor. The study was performed for the small and medium sized reactor and the large sized reactor by focusing on the development of the analysis technology for the passive components. Among the identified concepts the once-through steam generator, the natural circulation of the integral reactor, heat pipe for containment cooling, and hydraulic valve were selected as the high priority items to be developed and the related studies are being performed for these items. For the large sized passive reactor, the study plans to extend the applicability of the best estimate computer code RELAP5/MOD3 which is widely used for the safety analyses of the reactor system. The improvement and supplementation study of the analysis modeling and the methodology is planned to be carried out for these purpose. The newly developed technologies are expected to be applied to the domestic advanced reactor design and analysis and these technologies will play a key role in extending the domestic nuclear base technology and consolidating self-reliance in the essential nuclear technology. 72 figs, 15 tabs, 124 refs. (Author).

  15. Advanced PWR technology development -Development of advanced PWR system analysis technology-

    International Nuclear Information System (INIS)

    The primary scope of this study is to establish the analysis technology for the advanced reactor designed on the basis of the passive and inherent safety concepts. This study is extended to the application of these technology to the safety analysis of the passive reactor. The study was performed for the small and medium sized reactor and the large sized reactor by focusing on the development of the analysis technology for the passive components. Among the identified concepts the once-through steam generator, the natural circulation of the integral reactor, heat pipe for containment cooling, and hydraulic valve were selected as the high priority items to be developed and the related studies are being performed for these items. For the large sized passive reactor, the study plans to extend the applicability of the best estimate computer code RELAP5/MOD3 which is widely used for the safety analyses of the reactor system. The improvement and supplementation study of the analysis modeling and the methodology is planned to be carried out for these purpose. The newly developed technologies are expected to be applied to the domestic advanced reactor design and analysis and these technologies will play a key role in extending the domestic nuclear base technology and consolidating self-reliance in the essential nuclear technology. 72 figs, 15 tabs, 124 refs. (Author)

  16. Effects on radial core power profile on core thermo-hydraulic behavior during reflood phase in PWR-LOCAs

    International Nuclear Information System (INIS)

    An investigation of the effects of the radial core power profile on the thermo-hydraulic behavior during the reflood phase in a PWR-LOCA has been conducted with the Slab Core Test Facility (SCTF). Since the power in an actual PWR is lower in the peripheral bundles than in the central bundles, the so called chimney effects due to radial core power profile are expected to improve the cooling of the higher power bundles. The SCTF simulates a full radius slab section of a PWR and therefore the effects of radial core power profile can be investigated. The revealed results of four forced-feed reflood tests in the SCTF are; (1) even with different radial core power profiles, flat distribution of the collapsed water level in the core are obtained for each test; (2) in the highest power bundle under the same total core power, steeper radial power profile gives higher heat transfer coefficient; and (3) redistribution of flow or cross flow between bundles is considered to be a major reason for the results described above. (author)

  17. Review of the high conversion-type core study. Review about PWR

    International Nuclear Information System (INIS)

    The study of the high conversion light water reactor was proposed by Edlund in the United States of America in 1975. This theme was aggressively studied in the 1980s. As the reason, the increase of Pu produced from the reprocessing and the delay of practical use of FBR is given. A high converter core can realize comparatively easily by changing core of existing PWR to the tight and short core. In this report, the high converter core study about PWR were reviewed. In the United States of America, the study has already ended but an aggressive study is carried forward in Germany, Japan and so on. In addition to the reactor physics computation, the wide range study of such as critical experiment, conduct experiment of the heat transfer, the fracture behavior of fuel and the reactor-type strategy are carried forward. To investigate these studies is extremely useful in examining a future. (author)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  1. IASCC evaluation of core barrel weld line in PWR

    International Nuclear Information System (INIS)

    There is a concern of IASCC (Irradiation Assisted Stress Corrosion Cracking) in Core barrel weld line in long term operation because it is located near the core region and exposed to high neutron flux and there is weld-induced residual stress. In this report, weld-induced residual stress analysis method for EBW (Electron Beam Welding) is validated by comparing calculated residual stress with measured stress in a mock-up. Using this analysis method, stress of Core barrel EBW weld line in operating condition is calculated and it is confirmed that the probability of IASCC initiation in weld line of Core barrel is considered to be low. (authors)

  2. A Tight Lattice, Epithermal Core Design for the Integral PWR

    International Nuclear Information System (INIS)

    An 8-year core design for an epithermal, water-cooled reactor has been developed based upon assessments of nuclear reactor physics, thermal-hydraulics and economics. An integral vessel configuration is adopted and self-supporting wire-wrap fuel is employed for the tight lattice of the epithermal core. A streaming path is incorporated in each assembly to ensure a negative void coefficient. A whole-core MCNP simulation of the tight core shows a negative void coefficient for any burnup with positive KEFF. The VIPRETM code has been used to calculate the critical heat flux (CHF) by means of an appropriate wire-wrap CHF correlation, specifically introduced in the source code. Economically, the high fuel enrichment (14% w/o 235U) and the very long core life (8 ys) lead to high lifetime-levelized unit fuel cycle cost (in mills/kWhre). However, both operation and maintenance and capital-related expenditures strongly benefited from the higher electric output per unit volume, which yielded quite small lifetime-levelized unit capital and operation and maintenance costs for the overall plant. Financing costs are included and an estimate is provided for the total lifetime-levelized unit cost of the epithermal core, which is about 20% lower than that of a more open lattice thermal spectrum core fitting into the same core envelope and with 4-year lifetime. (authors)

  3. Core design study for power uprating of integral primary system PWR

    International Nuclear Information System (INIS)

    Highlights: • We propose a 20% power uprate of an integral primary system PWR for a better economic performance. • The power uprate is achieved mainly by optimizing core radial power peaking. • By enrichment zoning, power peaking of the proposed core is found to be around 1.43. • The predicted steady state MDNBR of the proposed core is found to be 3.454. • The results show that the proposed core design satisfies all design criteria defined in this study. - Abstract: Core design study for power uprating of integral primary system PWR has been performed. The selected reference core design is a four-year straight burn option of the International Reactor Innovative and Secure (IRIS). The objective of this study is to evaluate the possibility of increasing thermal power of the IRIS reactor by 20%, for a better economic performance. This study includes core neutronic and thermal hydraulic aspects. The power uprate is achieved by minimizing core radial power peaking, and by increasing fuel enrichment and coolant mass flow rate. Parametric calculations have been carried out to optimize the power-uprated core design, with the objective of obtaining relatively low core power peaking and similar initial reactivity with that of the reference core. The optimization is achieved by means of enrichment zoning and application of burnable poison, with different configuration from that of the reference core. The performance of power-uprated core is compared with the reference core. The calculation results show that the proposed core-uprated design with thermal power of 1200 MWt shows preferable characteristics, such as low power peaking of around 1.43, negative reactivity coefficients, and relatively high MDNBR of 3.454

  4. In-core detector activation rate for a PWR assembly

    International Nuclear Information System (INIS)

    The in-core detector system is the principal source of information for determining relative assembly powers, and maximum fuel rod powers in a reactor core. The detector signals are used in conjunction with pre-calculated factors, and appropriate normalizations, to obtain measured power values. Considerable reliance is placed on the accuracy of in-core detector inferred power distributions in reactor operations, and in the verification of calculational methods. The objective of this study was to compare results from standard design codes for the in-core detector activation rate (and the fission rate distribution in an assembly), to results obtained from a detailed calculation performed with a continuous energy Monte Carlo program with ENDF/B-V nuclear data

  5. Some factors affecting radiative heat transport in PWR cores

    International Nuclear Information System (INIS)

    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 UO2. 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 UO2 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 ∼ 22000C 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)

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

    International Nuclear Information System (INIS)

    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

  7. Development and validation of the 3-D PWR core dynamics SIMTRAN code

    International Nuclear Information System (INIS)

    We discuss the main features and results of the SIMTRAN development and validation work. Included in the first are the extension of the nodal neutronic solution to account for intranodal shape and spectrum, due to both heterogeneities and flux gradients, the implicit scheme for spatial kinetics with six delayed neutron precursors and the integration of the neutronic and thermohydraulic solutions on an staggered time mesh. Validation results are discussed for the NEACRP 3-D PWR Core Transient Benchmark and an actual transient with sudden increase of core flow occurred in the Vandellos-II 3-loop PWR NPP. Agreement with the reference numerical solution and measured plant data is shown for both problems. (orig./DG)

  8. An expert system for PWR core operation management

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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 (UO2 - PuO2) fuel assemblies up to 50% of the core, together with UO2 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 UO2 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 UO2. (author)

  11. Advanced ion exchange resins for PWR condensate polishing

    International Nuclear Information System (INIS)

    The severe chemical and mechanical requirements of a pressurized water reactor (PWR) condensate polishing plant (CPP) present a major challenge to the design of ion exchange resins. This paper describes the development and initial operating experience of improved cation and anion exchange resins that were specifically designed to meet PWR CPP needs. Although this paper focuses specifically on the ion exchange resins and their role in plant performance, it is also recognized and acknowledged that excellent mechanical design and operation of the CPP system are equally essential to obtaining good results. (authors)

  12. Engineering design feasibility of low boron concentration core in PWR

    International Nuclear Information System (INIS)

    In pressurized water reactor operation, higher level of soluble boron concentration could contribute higher impact from boron dilution situations, higher amount of liquid waste, and higher radiation dose to operators from higher corrosion potential to cladding and structure. Two practical and feasible means to reduce the maximum boron concentration were investigated in this study. A technically straightforward, possible means, can be achieved either by implementation of enriched boric acid (Eba) or by increasing more shim rod (fixed burnable absorber) worth. A simplest option is that the Eba is applied into reference core (Ref) design, OPR-1000 design, Ulchin unit-5 by allowing use of same fuel assemblies and core design without changing any nuclear design methodology used in that Ref design. Although results of Eba option proved its favorable power distribution and peaking factor, its moderator temperature coefficient (MTC) value reached positive, 3.25 pcm/ C at 40 EFPD which is beyond the design safety limit. An alternative option with more shim rods in fuel assemblies was tried with four types of integral burnable absorbers: gadolinia, integral fuel burnable absorber (Ifba), erbium and alumina boron carbide. Four core design candidates have been developed by keeping major engineering designs and preserving equivalent fuel enrichment level used in Ref design. However, all optimal designs were targeted to achieve comparable discharge burnup as well as favorable design safety parameters. The comparative analysis between Ref and optimal core designs is presented here. One of them is suggested as the most promising and favorable low boron core (Lbc) design in this framework. The proper combination of axial and radial enrichment zoning pattern in Lbc design candidate with Ifba-bearing fuel assemblies at equilibrium cycle, could bring 2 times narrower axial offset variation than that of Ref design, and maintain acceptable power peaking factor around 23% lower than

  13. Hydraulic test for non-instrumented capsule of advanced PWR fuel pellet

    International Nuclear Information System (INIS)

    This report presents the results of pressure drop test, vibration test and endurance test for Non-instrumented Capsule of Advanced PWR Fuel Pellet which were designed fabricated by KAERI. From the pressure drop test results, it is noted that the flow rate across the Non-instrumented Capsule of Advanced PWR Fuel Pellet corresponding to the pressure drop of 200 kPa is measured to be about 7.45 kg/sec. Vibration frequency for the Non-instrumented Capsule of Advanced PWR Fuel Pellet ranges from 13.0 to 32.3 Hz. RMS(Root Mean Square) displacement for the fuel rig is less than 11.6 μm, and the maximum displacement is less than 30.5 μm. The endurance test was carried out for 103 days and 17 hours

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

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H. M.; Energy Technology

    2006-01-31

    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

  15. PWR core safety analysis with 3-dimensional methods

    International Nuclear Information System (INIS)

    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

  16. Core liquid level depression due to manometric effect during PWR small break LOCA

    International Nuclear Information System (INIS)

    In the previous study, it is reported that the core collapsed liquid level was depressed nearly to the core bottom and the dryout of the core was observed in the early stage of the PWR cold leg small break loss-of-coolant accident (LOCA) experiment. The manometric effect due to the liquid seal formation in the loop seal and the difference of the liquid holdup between the steam generator (SG) upflow-side and downflow-side caused a depression of the core collapsed liquid level. The core liquid level was recovered just after the loop seal was cleared. The bypass between the core side and the downcomer side affects the core liquid depression. Four 5% cold leg break experiments with the different core bypass location, configuration and size were conducted to clarify the bypass effect. When the bypass was relatively small (less than 3% bypass of the initial core flow before the break), the timing of the loop seal clearing delayed with the bypass. When the bypass was relatively large (9.2% of the core flow), the loop seal clearing took place after the break uncovery and the timing was significantly delayed. In general, the smaller minimum core collapsed liquid level was obtained at the earlier timing of loop seal clearing due to the smaller bypass. (author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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)

  19. Solution of the stationary state of the PWR MOX/UO-2 core transient benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Seubert, A.; Langenbuch, S.; Zwermann, W. [Gesellschaft fuer Anlagen- und Reaktorsicherheit GRS mbH, Forschungsinstitute, D-85748 Garching (Germany)

    2006-07-01

    The multi-group Discrete Ordinates transport code DORT is applied to solve the stationary state of the OECD/NEA PWR MOX/UO-2 Core Transient Benchmark. Pin cell homogenised cross sections in 16 energy groups and P{sub 1} scattering order have been obtained by fuel assembly burn-up calculations using HELIOS. In this paper, we report on the details of our calculations for this benchmark problem and show our results to be in good agreement with an MCNP Monte Carlo solution with nuclear point data and a multi-group DeCART Method of Characteristics solution. (authors)

  20. Solution of the stationary state of the PWR MOX/UO-2 core transient benchmark

    International Nuclear Information System (INIS)

    The multi-group Discrete Ordinates transport code DORT is applied to solve the stationary state of the OECD/NEA PWR MOX/UO-2 Core Transient Benchmark. Pin cell homogenised cross sections in 16 energy groups and P1 scattering order have been obtained by fuel assembly burn-up calculations using HELIOS. In this paper, we report on the details of our calculations for this benchmark problem and show our results to be in good agreement with an MCNP Monte Carlo solution with nuclear point data and a multi-group DeCART Method of Characteristics solution. (authors)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  4. Advanced PWR fuel assembly development programs in Korea

    International Nuclear Information System (INIS)

    Both KNFC and Westinghouse have continued to focus on developing products that will meet the challenge of increasing fuel duty requirements in Korea. These higher duty conditions include higher energy core designs through improved plant capacity factors, power uprate, extended fuel burnup, peaking factor increases, and more severe coolant chemistry (including high lithium concentration). Recent advanced fuel development activities in Korea include implementation of the 17x17 Robust Fuel Assembly (RFA), which is currently in operation with excellent performance in the United States and Europe, as well as the 16x16 PLUS7TM fuel assembly for use in KSNP plants. KNFC and Westinghouse are jointly developing advanced fuel that will meet future fuel duty challenges of 17x17 and 16x16 Westinghouse type plants. This paper focuses on advanced fuel assembly development programs that are underway and how these designs demonstrate improved margins under high duty plant operating conditions. In designing for these high duty conditions key design considerations for the various operational modes (i.e. power uprating, high burnup, long cycles, etc.) must be identified. These design considerations will include the traditional factors such as safety margin (DNB and LOCA), fuel rod design margin (e.g. corrosion, internal pressure, etc.) and mechanical design margins, among others. In addressing these design considerations, the fundamental approach is to provide additional design margin through materials, mechanical, and thermal performance enhancements, to assure flawless fuel performance. The foundation of all fuel designs is the product development process used to meet the demands of modern high duty operation including power uprating, high burnup, longer cycles, and high-lithium coolant chemistries. These advanced fuel assembly designs incorporate features that provide improved mechanical design margin, as well as thermal performance margin (DNB). Enhanced grid designs result in a

  5. Modeling and simulation of the core in PWR nuclear power plant based on PAnySimu

    International Nuclear Information System (INIS)

    Modeling and simulation analysis on the core of PWR nuclear power plant based on PAnySimu Simulation Support System. We have divided the core into five models by studying the Unit 3/4's actual core structure of Ling Ao Phase Ⅱ, which are, power calculation, calculation of core transmission, control rod reactions, reactivity feedback calculation and poison calculation. On this basis, analyses the core neutron flux, consider the influence of the control rod position, fuel and moderator temperature, xenon and samarium poisoning, boron concentration on neutron Biomass. The various modules of algorithms and primitives are defined by using PAnySimu simulation system. Then prepare the corresponding module program in the C++ Builder 5.0 environment, and storage and debug algorithm. After running the module, build core model simulation system in accordance with logical relations. At last do the dynamic simulation separately on the reaction of individual modules and systems under other circumstances of disturbance. The analysis of the real-time dates shows that results are reasonable. For some hard core within the experimental operation in the other, the simulation method to obtain experimental data, is of great significance. (authors)

  6. Studies on the flow distribution in the PWR reactor core using ANSYS CFX

    International Nuclear Information System (INIS)

    The validated and verified ANSYS-CFX model was used to calculate the flow and pressure distribution in the reactor pressure vessel of a four-loop PWR plant for the condition zero load (50 C) during operation with all four primary coolant pumps. Due to the strong flow homogenization in the lower core half and the transfer of the main flow in the upper half toward the edge of the core the ANSYS CFX calculation shows that the local peaking of the pressure loss or the fuel element flow force with respect to the mean value of 7% is significantly lower than the actually used margin for fuel element holding considering the inhomogeneity (14%). The comparison of the calculated results with the measured data using the 1:10 model confirmed that the ANSYS CDX results cannot be simply transferred to PWR plants with flow skirt. For such plants specific CDF analyses with an appropriate Model for the flow skirt in the ROPV would be necessary.

  7. Development of enriched Gd-155 and Gd-157 burnable poison designs for a PWR core

    International Nuclear Information System (INIS)

    In this study, a genetic algorithm developed by the authors was applied to design the optimal enriched Gd-155 and Gd-157 burnable poisons in a reference PWR TMI-1 core. The CASMO-4/TABLES/SIMULATE-3 package calculated the neutronic performance of the enriched UO2/Gd2O3 fuel pin configurations. These configurations included different fractions of neutron absorbing isotopes Gd-155 and Gd-157, and 100 w/o enriched Gd-155 designs. Fuel cost analysis was performed to evaluate the economical benefits of these optimized enriched gadolinium designs. The break-even point for unit Gd-155 enrichment cost was determined to be around ∼$30/gram-Gd-155 with current unit cost scenario. The projected savings were 3.13% in gross and 2.08% in net compared to total fuel cycle cost of a reference TMI-1 core loading, if all of the 68 feed assemblies would be replaced with the optimized designs

  8. Examination of offsite radiological emergency measures for nuclear reactor accidents involving core melt. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Aldrich, D.C.; McGrath, P.E.; Rasmussen, N.C.

    1978-06-01

    Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each measure were developed and are discussed. Potential PWR core-melt radioactive material releases are separated into two categories, ''Melt-through'' and ''Atmospheric,'' based upon the mode of containment failure. Protective measures are examined and compared for each category in terms of projected doses to the whole body and thyroid. Measures for ''Atmospheric'' accidents are also examined in terms of their influence on the occurrence of public health effects.

  9. Flexibility control and simulation with multi-model and LQG/LTR design for PWR core load following operation

    International Nuclear Information System (INIS)

    Highlights: ► The nonlinear model and linear multi-model of a PWR core are developed. ► The LQG/LTR robust control is used to design local controllers of the core. ► LTR principles are analyzed and proved theoretically. ► Flexibility control is proposed to design flexibility controllers for the core. ► The nonlinear core load following control system is effective. - Abstract: The objective of this investigation is to design a nonlinear Pressurized Water Reactor (PWR) core load following control system. On the basis of modeling a nonlinear PWR core, linearized models of the core at five power levels are chosen as local models of the core to substitute the nonlinear core model in the global range of power level. The Linear Quadratic Gaussian with Loop Transfer Recovery (LQG/LTR) robust optimal control is used to contrive a controller with the robustness of a core local model as a local controller of the nonlinear core. Meanwhile, LTR principles are analyzed and proved theoretically by adopting the matrix inversion lemma. Based on the local controllers, the principle of flexibility control is presented to design a flexibility controller of the nonlinear core at a random power level. A nonlinear core model and a flexibility controller at a random power level compose a core load following control subsystem. The combination of core load following control subsystems at all power levels is the core load following control system. Finally, the core load following control system is simulated and the simulation results show that the control system is effective

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

    International Nuclear Information System (INIS)

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

  11. Development of an advanced man-machine system for Japanese PWR plants

    International Nuclear Information System (INIS)

    Advanced Man Machine System for Japanese PWR plants (MMS - PWR) is a prototype system of the operator supporting system that has been developed by Mitsubishi Heavy Industries, Ltd., Mitsubishi Electric Co. and Mitsubishi Atomic Power Industries, Inc. for five years from 1987 to 1991, under the financial support of MITI (the Ministry of International Trade and Industry of Japanese government). The aim of this system development is to further increase operation reliability and operability of PWR plants. For this purpose the knowledge engineering and the up-to-date computer technologies have been introduced into the design of a prototype system that can offer the information, infer and judge according to operator's thinking process in grasping the plant status and the operations. Also the system has been verified. A prototype system has the following supporting functions: (1) Operator Supporting Function for Normal Operation: Flexible operator supporting function for re-start-up and load following operation that will be needed for Japanese PWR plants in the future. (2) Operator Supporting Function for Abnormalities and Accidents: Early detection and identification of abnormality or accident of the plant, and guidance to appropriate countermeasures. (3) Operator Supporting Function for Maintenance: Supporting function for evaluation of the influence of maintenance on plant components, as well as for the isolation and restoration procedures during plant operation. (4) Optimum Operation Surveillance Function: Intelligence man machine interface that enables operators to understand various plant data precisely and offers the proper answers of what they need to know. (author). 6 figs, 2 refs

  12. Remarks on methods of evaluation of aerosol sources related to PWR core meltdown accidents

    International Nuclear Information System (INIS)

    The paper tries to demonstrate the conceptional background of the KfK core melting program, which has been started in 1973, and which is scheduled to be terminated by 1986. The paper also summarizes the main findings of the SASCHA program, with the aid of which the enveloping fission product release from the primary system into the containment during a PWR core melt accident has been investigated. The fractions of release from the fuel determined in the experiment are undoubtedly in the range of 70% to 100% for the radiologically most important elements I, Cs, Te. The reduction in release from the primary circuit due to deposition is 50% at the maximum. A considerable portion resuspended must be deducted from that value. The retention of iodine and aerosol particles in the safety containment amounts to several orders of magnitude (up to 5). Likewise, the decrease in the population dose by spread and dilution in the environment and due to other parameters attains several orders of magnitude (up to 7). Consequently, particle retention by a factor of 2 or 3 in the primary circuit is negligible. - Our present knowledge is completely satisfactory for analyzing the so-called source term in core melt accidents. The wish to develop more detailed codes related to core degradation and to activity release from the primary circuit has many understandable causes. However, there is no single technical reason in favor of spending much money in order to materialize this wish. (orig./HP)

  13. Feasibility study for core cooling performance using SG secondary-side depressurization in PWR

    International Nuclear Information System (INIS)

    In light of the lessons learned from station blackout accidents of the Fukushima Dai-ichi reactor, it is important to line up various cooling measures for reactor core and containment. We are progressing to develop a reliable alternative safety measure to cool the reactor core under small break loss-of-coolant accident (SBLOCA) of PWR using SG secondary-side depressurization. In this research, we aim to promote an early activation of accumulators (ACC) and low-pressure injection (LPI) system to assure the core cooling by an early SG secondary-side depressurization even under loss of core cooling functions by high-pressure injection system. The feasibility study of the safety measure then is being performed by the ROSA / large-scale test facility (LSTF), where tests can be conducted under full-pressure, at Japan Atomic Energy Agency since 2011. The applicability of safety evaluation code M-RELAP5 is also being investigated to establish an evaluation technique for an actual reactor. In this paper, we will present the outline of the safety measure, typical test results and M-RELAP5 calculation results. It is confirmed that the new safety measure is feasible and M-RELAP5 can apply to the SBLOCA transients. (author)

  14. Modeling and design of a reload PWR core for a 48-month fuel cycle

    International Nuclear Information System (INIS)

    The objective of this research was to use state-of-the-art nuclear and fuel performance packages to evaluate the feasibility and costs of a 48 calendar month core in existing pressurized water reactor (PWR) designs, considering the full range of practical design and economic considerations. The driving force behind this research is the desire to make nuclear power more economically competitive with fossil fuel options by expanding the scope for achievement of higher capacity factors. Using CASMO/SIMULATE, a core design with fuel enriched to 7w/o U235 for a single batch loaded, 48-month fuel cycle has been developed. This core achieves an ultra-long cycle length without exceeding current fuel burnup limits. The design uses two different types of burnable poisons. Gadolinium in the form of gadolinium oxide (Gd2O3) mixed with the UO2 of selected pins is sued to hold down initial reactivity and to control flux peaking throughout the life of the core. A zirconium di-boride (ZrB2) integral fuel burnable absorber (IFBA) coating on the Gd2O3-UO2 fuel pellets is added to reduce the critical soluble boron concentration in the reactor coolant to within acceptable limits. Fuel performance issues of concern to this design are also outlined and areas which will require further research are highlighted

  15. An evaluation of the failure of a PWR lower head during a core meltdown accident

    International Nuclear Information System (INIS)

    This paper presents an analysis of the failure of lower vessel head during a core meltdown accident. The analysis is limited to PWR systems with no penetration tubes attached to the lower vessel head. The case considered is characterized by a small quantity of corium and a relatively slow discharge into the lower plenum. The assumption of the breakup of the jet stream results in the solidification of debris particles and the formation of a debris bed thermally attacking the lower head wall. Detailed analyses were performed to determine the debris/water interaction, ablation of the lower head wall, and the time of vessel failure. Parameters which have significant effect on the results were identified. Parametric studies were performed to reflect uncertainties associated with the various phenomenological processes occurring during corium relocation into the lower head

  16. 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...

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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 9th National Conference on Mechanics, Ha Noi, December 8-9, 2012, the 10th 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

  19. CFD simulation of fibre material transport in a PWR core under loss of coolant conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, Thomas; Grahn, Alexander; Kliem, Soeren; Weiss, Frank-Peter [Forschungszentrum Dresden-Rossendorf e.V., Dresden (Germany). Inst. of Safety Research

    2010-05-15

    The aim of the numerical simulations carried out in this study was to determine how and where mineral wool fibres transported to the core by ECC water during a LOCA are deposited across the grid spacers of the fuel elements of a German PWR. The spacer grid is modelled as a strainer which completely retains the insulation material carried by the coolant and reaching the plane of the spacers. The accumulation of the insulation material gives rise to the formation of a compressible fibrous cake whose permeability to the coolant flow is calculated in terms of the local amount of deposited material and the local value of the superficial liquid velocity. The calculations showed that the fiber material at the uppermost spacer grid plane is not evenly distributed. First, it is accumulated at the positions of the break-through channels. Later when the inner circulation in the core has stopped, the insulation material can also be distributed into other regions of the spacer plane. Further investigations are necessary to determine the accumulation of insulation material for a longer period of time. Also steam production in the core or re-suspension of the insulation material during back flow should be considered. Moreover, the geometry modeling should be improved taking into account the real structures in the upper plenum and the geometry of the ECC injection nozzle ('Hutze'). (orig.)

  20. Nodal parametrization of cross sections for a 3D PWR core nodal model

    International Nuclear Information System (INIS)

    Having an accurate PWR core analysis system as SEANAP, requires continuous developments in order to improve the internal calculation models. At present a new 3D nodal neutronics model is being developed, for future implementation in the coupled neutronic-thermalhydraulic SIMTRAN code, operating on-line at several Spanish nuclear power plants. This development has the goal of improving the 3D effect treatments in SIMTRAN, taking into account all the heterogeneities in detail. The new neutronics model includes for the nodal treatment the same methodology of cross section parametrization used in the 2D core DELFOS code. A 2 group extended library per node type is generated likewise as the cell library, including all the partial derivatives with local variable and the spectral history and neighborhood indexes. Thus, the cross sections per node are determined with a set of independent, local and generalized corrections. In this way the neutronics database preparation for the core simulator is simplified. The purpose of this paper is to discuss the cross section nodal parametrization for the treatment of heterogeneity in the 3D formulation. (Author) 5 refs

  1. 3D thermal-hydraulic analysis on core of PWR nuclear power station

    International Nuclear Information System (INIS)

    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

  2. CFD simulation of fibre material transport in a PWR core under loss of coolant conditions

    International Nuclear Information System (INIS)

    The aim of the numerical simulations carried out in this study was to determine how and where mineral wool fibres transported to the core by ECC water during a LOCA are deposited across the grid spacers of the fuel elements of a German PWR. The spacer grid is modelled as a strainer which completely retains the insulation material carried by the coolant and reaching the plane of the spacers. The accumulation of the insulation material gives rise to the formation of a compressible fibrous cake whose permeability to the coolant flow is calculated in terms of the local amount of deposited material and the local value of the superficial liquid velocity. The calculations showed that the fiber material at the uppermost spacer grid plane is not evenly distributed. First, it is accumulated at the positions of the break-through channels. Later when the inner circulation in the core has stopped, the insulation material can also be distributed into other regions of the spacer plane. Further investigations are necessary to determine the accumulation of insulation material for a longer period of time. Also steam production in the core or re-suspension of the insulation material during back flow should be considered. Moreover, the geometry modeling should be improved taking into account the real structures in the upper plenum and the geometry of the ECC injection nozzle ('Hutze'). (orig.)

  3. 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)

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

    International Nuclear Information System (INIS)

    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)

  5. Development and validation process of the advanced main control board for next Japanese PWR plants

    International Nuclear Information System (INIS)

    The purpose of main control room improvement is to reduce operator workload and potential human errors by offering a better working environment where operators can maximize their abilities. Japanese pressurized water reactor (PWR) utilities and Mitsubishi group have developed a touch -screen-based main control console (i.e. advanced main control room) the next generation PWRs to further improve the plant operability using a state of the art electronics technology. The advanced main control room consists of an operator console, a supervisor console and large display panels. The functional specifications were evaluated by utility operators using a prototype main control console connected to a plant simulator. (author)

  6. Advanced core monitoring technology for WWER reactors

    International Nuclear Information System (INIS)

    The Westinghouse BEACON online monitoring system has been developed to provide continuous core monitoring and operational support for pressurized water reactor using movable detectors (fission chamber) and core thermocouples. The basic BEACON core monitoring methodology is described. Traditional WWER reactors use rhodium fixed in-core detectors as the means to provide detailed core power distribution for surveillance purposes. An adapted version of the BEACON advanced core monitoring and support system is described which seems to be, due to the different demand/response requirements, the optimal solution (for routine surveillance and anomaly detection) for WWER reactors with existing fixed in-core detectors. (Z.S.) 4 refs

  7. Investigations of boron transport in a PWR core with COBAYA3/SUBCHANFLOW inside the NURESIM platform

    International Nuclear Information System (INIS)

    Highlights: • Simulation of boron slug transport in a core with coupled codes. • Spatial and temporal discretization to describe boron dilution scenarios. • Code validation by code-to-code comparison. • Boron dilution simulation including cross flow between fuel assemblies. - Abstract: Multi-scale, multi-physics problems reveal significant challenges while dealing with coupled neutronic/thermal-hydraulic solutions. Current generations of reactor dynamic codes applied to Light Water Reactors (LWRs), and in the context of this paper, pressurized water reactors (PWRs), are based on 3D neutronic nodal methods coupled with single or two phase flow thermal hydraulic system or sub-channel codes. This paper describes the extension of the coupling scheme between the 3D neutron diffusion codes COBAYA3 and DYN3D, and the sub-channel thermal-hydraulic code SUBCHANFLOW for the simulation of boron dilution transients. This includes the validation of the boron transport model of SUBCHANFLOW. The coupling of COBAYA3 and DYN3D with SUBCHANFLOW is performed inside the NURESIM platform making use of the novel automatic mesh superposition for code coupling. Boron transport models are needed for the simulation of boron dilution transients following a SBLOCA (after loop-seal clearing). In this case, the mixing is a key mechanism determining the positive reactivity insertion in the core. The results obtained with the coupled codes COBAYA3/SUBCHANFLOW, DYN3D/SUBCHANFLOW will be presented and discussed for two transient scenarios; a homogeneous and a heterogeneous boron dilution problem for a mini-core and for a PWR core defined in the NURISP boron dilution Benchmark

  8. Understanding using the Haling Power Depletion (HPD) as a guide for designing PWR cores

    International Nuclear Information System (INIS)

    Highlights: ► The Haling Power Depletion (HPD) may be used as a guide to design low-leakage cores. ► The paper presents studies that were performed to better understand the HPD. ► The first phase covered solution to design low leakage core for a given cycle length. ► Techniques using the HPD results are developed to guide placing of BPs in the core. - Abstract: The Pennsylvania State University (PSU) is using the university version of the Studsvik Scandpower Code System (CMS) for research and education purposes. Preparations have been made to incorporate the CMS codes into the PSU Nuclear Engineering graduate class “Nuclear Fuel Management” course. The information presented in this paper has been developed during two phases of preparation of the material for the course. In the first phase, the Haling Power Depletion (HPD) was presented in the course for the first time. The HPD method has been criticized as not valid by many in the field even though it has been successfully applied at PSU for the past 20 years. It was noticed during the first phase that the radial power distribution (RPD) for low leakage cores during depletion remained similar to that of the HPD during most of the cycle and have close maximum normalized power values (NPmax’s) and cycle lengths. Thus, the Haling Power Depletion (HPD) may be used as a guide for conveniently designing mainly low leakage PWR cores because the HPD and actual core are similar. Studies were then made to better understand the HPD. Many different core configurations can be computed quickly with the HPD without using Burnable Poisons (BPs) to produce several excellent low leakage core configurations that are viable for power production. The first phase covered the solution to design a low leakage core for a cycle length of 16 GWD/MTU. Once the HPD core configuration is chosen as a potential core, it is followed by establishing a BP design to prevent violating any of the safety constraints during depletion. The

  9. Advanced fuel developments to improve fuel cycle cost in PWR

    International Nuclear Information System (INIS)

    Increasingly lower fuel cycle costs and higher plant availability factors have been two crucial components in keeping the overall cost of electricity produced by nuclear low and competitive with respect to other energy sources. The continuous quest to reduce fuel cycle cost has resulted in some consolidated trends in LWR fuel management schemes: smaller number of feed fuel assemblies with longer residence time; longer cycles, with 18-month cycle as the predominant option, and some plants already operating on, or considering, 24-month refueling intervals; higher power ratings with many plants undergoing power uprates. In order to maintain or improve fuel utilization for the longer cycles and/or higher power ratings, the licensed limits in fuel fissile content (5.0 w/o U235 enrichment) and discharge burnup (62 GWd/tHM for the peak pin) have been approached. In addition, Zr-based fuel cladding materials are also being challenged by the resulting increased duty. For the above reasons further improvements in fuel cycle cost have to overcome one or more of the current limits. This paper discusses an option to break through this 'stalemate', i.e. uranium nitride (UN) fuel with SiC clad. In UN the higher density of the nitride with respect to the oxide fuel leads to higher fissile content and reduction in the number of feed assemblies, improved fuel utilization and potentially higher specific powers. The SiC clad, among other benefits, enables higher clad irradiation, thereby exploiting the full potential of UN fuel. An alternative to employing UN fuel is to maintain UO2 fuel but boost the fissile content increasing the U235 enrichment beyond the 5 w/o limit. The paper describes and compares the potential benefits on fuel cycle cost of either option using realistic full-core calculations and ensuing economic analysis performed using Westinghouse in-house reactor physics tools and methodologies. (author)

  10. The possibility of building nuclear power plant free from severe accident risk: PWR NPP with advanced all passive safety cooling systems (AAP SCS)

    International Nuclear Information System (INIS)

    A complete set of advanced all passive safety cooling systems (AAP SCS) for PWR NPP, actuated by natural force has been put forward in the article. Here the natural force mainly means the fore, which created by change of pressure distribution in the first loop of PWR as a result of operational regime conversion from one to another, including occurrence of accident situation. Correspondent safety cooling system will be actuated naturally and then put it into passive operation after occurring some kind of accident, so accidental situation will be mitigated right after it's occurrence and core residual heat will be naturally moved from the active core to the ultimate heat sink. There is no need to rely on automatic control system, any active equipment and human actions in all working process of the AAP SCS, which can reduce the probability of severe accident to zero, so as to exclude the need of evacuation plan around AAP nuclear power plant and eliminate the public's concern and doubt about nuclear power safety. Implementation of the AAP SCS concept is only based on use of evolutionary measures and state-of-the-art technology. So at present time it can be used for design of new-type third generation PWR nuclear power plant without severe accident risk, and for modernization of existing second generation nuclear power plant. (authors)

  11. A PN-based approach along PSO scheme for PWR core reloading patterns optimization

    International Nuclear Information System (INIS)

    Highlights: ► PSO algorithm along PN improves optimum loading pattern. ► The method provides great geometrical flexibility. ► The method uses a well known optimization algorithm. ► There is a capability of using high order PN in the algorithm. - Abstract: Core performance analysis constitutes an essential phase in core fuel management optimization. The output consists mainly of the neutron flux and core power distributions which are needed for deriving the safety related thermal margins. Based on the results of the core simulation, feasible options of loading patterns and control strategies can be specified within the broader scope. In other words, the output is employed in the logical decisions where the fuel management strategies are determined over a few core cycles. This work is focused on the strategy for obtaining a core reload optimization based on using continuous version of particle swarm optimization, PSO, along PN approximation for core analysis. In the last decades the obstruction one faced to use Boltzmann equation for neutron transport in optimization algorithms, was the formidable requirement of computer storage and running time. Now days, however, development of computer storage and running time make it possible to implement advanced methods for particle transport. In this paper an even-parity spherical harmonic expansion for angular distribution and finite element discretization for space variable is used to treat core analysis of particle swarm optimization. It is illustrated that the increasing spherical expansion of angular flux improves the obtained core loading pattern through PSO algorithm.

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

    International Nuclear Information System (INIS)

    Fission product releases from the RCB associated with hypothetical core-melt accidents ABβ, S2CDβ 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

  13. Sensitivity and uncertainty calculations methods of neutronics parameters in PWR cores. Part I. Theory and sensitivity calculations

    International Nuclear Information System (INIS)

    Sensitivity and uncertainty calculations methods of neutronics parameters in pressurized light water reactors have been developed. The sensitivity is composed of three terms; the first is the sensitivity of cell-averaged multi-group cross-sections relative to multi-group infinite dilution cross-sections, the second is the sensitivity of assembly averaged few-group macroscopic cross-sections relative to cell-averaged multi-group cross-sections, and the third is the sensitivity of neutronics parameters in PWR cores relative to few-group macroscopic cross-sections. Combining the three sensitivities, the sensitivity of neutronics parameters in PWR cores relative to multi-group infinite dilution cross-sections is obtained. The discussion of this method will be presented in two papers; the present paper is part I, where the theory and some numerical results for typical pin cells, fuel assemblies and a simple PWR core are shown. The present method gives us multi-group sensitivities for individual nuclides in each reaction type, and wide ranges of applications are possible to the fields such as cross-section adjustment and uncertainty reduction. (author)

  14. Maintenance method for irradiation assisted stress corrosion cracking of PWR core internals

    International Nuclear Information System (INIS)

    Irradiation-assisted stress corrosion cracking of IASCC is a well-known result of age-related degradation of baffle former bolts which are an integral component of PWR core internals. However, methods for analyzing the causes and assessing IASCC have yet to be established and are being studied. The baffle former bolts are components of the baffle structure and in order to maintain the integrity of the baffle structure, an effective approach for maintaining the bolts must be provided. This paper proposes an effective method of maintaining the functions of the baffle structure by improving the configuration and/or material of baffle former bolts. A relaxed configuration of the bolt neck reduces the stress and can approximately halve the damage caused by IASCC. An improvement in the threshold value of materials vulnerability to IASCC would not only extend the service life of all bolts but could also substantially retard IASCC of the bolts installed at upper and lower ends where flux is low. If the threshold fluence to IASCC could be made higher than approximately 5x1025 n/m2 (E>0.1 MeV, 340degC), the bolts installed at lower and upper ends, which are conventionally replaced 30 years after plant start-up, would remain intact for 60 years after start-up. Hence, this approach is a very effective maintenance measure. (author)

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

    International Nuclear Information System (INIS)

    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 UO2 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)

  16. Validation of NuStar's PWR core analysis system

    International Nuclear Information System (INIS)

    A new PWR core analysis system has recently been developed in NuStar. It adopts not only conventional theoretical framework that widely used in today's production code, but also latest progress that achieved internationally and within the company. To validate the system, a total of 44 operation cycles of Qinshan Phase 1 and 2 reactors are evaluated and results are compared against the measurement data. Statistics analyses of prediction errors for both low-power startup physics test states and high-power normal operation conditions demonstrate that the development of the system is quite a success. It not only has a generic applicability for reactors with different design, but also has a very good accuracy for parameters that are measurable at the site. Among all the validated cases, only for very rare cases that the acceptance tolerance is violated, and the error distributions for all the validated parameters present to be good ones that are close to the normal distribution. Based on the validation results, the system is considered to be qualified for practical applications. (author)

  17. Advanced passive PWR AC-600: Development orientation of nuclear power reactors in China for the next century

    International Nuclear Information System (INIS)

    Based on Qinshan II Nuclear Power Plant that is designed and constructed by way of self-reliance, China has developed advanced passive PWR AC-600. The design concept of AC-600 not only takes the real situation of China into consideration, but also follows the developing trend of nuclear power in the world. The design of AC-600 has the following technical characteristics: Advanced reactor: 18-24 month fuel cycle, low neutron leakage, low power density of the core, no any penetration in the RPV below the level of the reactor coolant nozzles; Passive safety systems: passive emergency residual heat removal system, passive-active safety injection system, passive containment cooling system and main control room habitability system; System simplified and the number of components reduced; Digital I and C; Modular construction. AC-600 inherits the proven technology China has mastered and used in Qirtshan 11, and absorbs advanced international design concepts, but it also has a distinctive characteristic of bringing forth new ideas independently. It is suited to Chinese conditions and therefore is expected to become an orientation of nuclear power development by self-reliance in China for the next century. (author)

  18. Performance evaluation of PSO and GA in PWR core loading pattern optimization

    Energy Technology Data Exchange (ETDEWEB)

    Khoshahval, F., E-mail: f_khoshahval@sbu.ac.i [Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Velenjak, Tehran (Iran, Islamic Republic of); Minuchehr, H. [Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Velenjak, Tehran (Iran, Islamic Republic of); Zolfaghari, A., E-mail: a-zolfaghari@sbu.ac.i [Engineering Department, Shahid Beheshti University, G.C., P.O. Box 1983963113, Velenjak, Tehran (Iran, Islamic Republic of)

    2011-03-15

    Research highlights: The performance of both GA and PSO methods in optimizing of a PWR core are adequate. It seems GA arrives to its final parameter value in a fewer generation than the PSO. The computation time for GA is higher than PSO. The GA-2 and PSO-CFA algorithms perform better in comparison to GA-1 and PSO-IWA. - Abstract: The efficient operation and fuel management of PWRs are of utmost importance. Recently, genetic algorithm (GA) and particle swarm optimization (PSO) techniques have attracted considerable attention among various modern heuristic optimization techniques. GA is a powerful optimization technique, based upon the principles of natural selection and species evolution. GA is finding popularity as design tools because of its versatility, intuitiveness and ability to solve highly non-linear, mixed integer optimization problems. PSO refers to a relatively new family of algorithms and is mainly inspired by social behavior patterns of organisms that live within large group. This study addresses the application and performance comparison of PSO and GA optimization methods for nuclear fuel loading pattern problem. Flattening of power inside the reactor core of Bushehr nuclear power plant (WWER-1000 type) is chosen as an objective function to prove the validity of algorithms. In addition the performance of both optimization techniques in terms of convergence rate and computational time is compared. It is found that, from an evolutionary point of view, the performance of both GA and PSO is quite adequate. But, GA seems to arrive at its final parameter value in a fewer generations than the PSO. It is also noticed that, the computation time for implemented GA in this work is too high in comparison to PSO.

  19. Performance evaluation of PSO and GA in PWR core loading pattern optimization

    International Nuclear Information System (INIS)

    Research highlights: → The performance of both GA and PSO methods in optimizing of a PWR core are adequate. → It seems GA arrives to its final parameter value in a fewer generation than the PSO. → The computation time for GA is higher than PSO. → The GA-2 and PSO-CFA algorithms perform better in comparison to GA-1 and PSO-IWA. - Abstract: The efficient operation and fuel management of PWRs are of utmost importance. Recently, genetic algorithm (GA) and particle swarm optimization (PSO) techniques have attracted considerable attention among various modern heuristic optimization techniques. GA is a powerful optimization technique, based upon the principles of natural selection and species evolution. GA is finding popularity as design tools because of its versatility, intuitiveness and ability to solve highly non-linear, mixed integer optimization problems. PSO refers to a relatively new family of algorithms and is mainly inspired by social behavior patterns of organisms that live within large group. This study addresses the application and performance comparison of PSO and GA optimization methods for nuclear fuel loading pattern problem. Flattening of power inside the reactor core of Bushehr nuclear power plant (WWER-1000 type) is chosen as an objective function to prove the validity of algorithms. In addition the performance of both optimization techniques in terms of convergence rate and computational time is compared. It is found that, from an evolutionary point of view, the performance of both GA and PSO is quite adequate. But, GA seems to arrive at its final parameter value in a fewer generations than the PSO. It is also noticed that, the computation time for implemented GA in this work is too high in comparison to PSO.

  20. Multimodel-based power-level control with state-feedback and observer for load-follow PWR core

    International Nuclear Information System (INIS)

    Highlights: • The equilibrium manifold and nonlinearity measure of the core are proposed. • The linear multi-model of the core is built based on the core nonlinearity measure. • A new state feedback control is used to design local controllers of the core. • Flexibility partitioning of model and control is presented for the nonlinear core. • The global stability of the core load follow control is analyzed. - Abstract: The purpose of this investigation is that a nonlinear Pressurized Water Reactor (PWR) core load following control system is designed and the global stability of the system is analyzed theoretically. On the basis of modeling a nonlinear PWR core and proposing the equilibrium manifold and the nonlinearity measure of the core to calculate the distribution situation of the core nonlinearity measure in the entire range of power level, linearized models of the core at five power levels are chosen as local models of the core and the set of local models is used to substitute the nonlinear core model. The full-state feedback control with a full-order observer is utilized to design a controller with robustness of every local model, which is treated as a local controller of the nonlinear core. The Kalman filter is contrived as an observer with robustness and the state feedback design with robustness is implemented via the robust pole assignment method. With the local models and local controllers, the flexibility partitioning of model and control is presented to design a decent flexibility controller of the nonlinear core at a random power level. A nonlinear core model and a flexibility controller at a random power level compose a core load following control subsystem. The combination of core load following control subsystems at all power levels is the core load following control system. Two global stability theorems are deduced to define that the core load following control system is globally asymptotically stable within the whole range of power level

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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.

  3. Application of the Monte Carlo thermal design analysis to evaluate uncertainties of the PWR core using the THALES subchannel code

    International Nuclear Information System (INIS)

    In order to maintain the safety of the reactor core, the minimum DNBR (Departure from Nucleate Boiling Ratio) in the PWR (Pressurized-Water Reactor) core remains higher than the DNBR limit during Condition I and II events. Therefore, it is important to adequately evaluate the thermal performance of the PWR core. To realistically evaluate the relationship among the uncertainties and reduce the conservatism resulting from the unknown phenomena, the Monte Carlo method is being used in many areas requiring the statistical approach. Especially, the Monte Carlo method is drawing attention as the method for the evaluation of the thermal performance of the PWR core. For the best estimate evaluation of the uncertainties in the PWR core, KEPCO Nuclear Fuel (hereinafter KEPCO NF) has been developing the thermal design analysis based on the Monte Carlo method. For the Monte Carlo thermal design analysis, various studies are conducted as follows. To generate the Gaussian random numbers, Gaussian random number generators are investigated. In this paper, Box-Muller, Polar, GRAND, and Ziggurat method are briefly reviewed. The random numbers are generated on the basis of the nominal value and uncertainty of the parameter. If the normal distribution is acceptable at 5% significance level through the normality tests, the random numbers are used for the Monte Carlo thermal design analysis. Using the subchannel code THALES (Thermal Hydraulic AnaLyzer for Enhanced Simulation of core) developed by KEPCO NF, the subchannel analyses are carried out considering the core operating parameters randomized, and then DNBR distribution is derived. Finally, if the DNBR distribution is statistically combined with the uncertainties of the other parameters, the DNBRT distribution can be obtained. From the DNBRT distribution, the DNBR limit is determined to avoid DNB (Departure from Nucleate Boiling) at a 95% probability at a 95% confidence level. Through the example calculation, it is verified that

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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, particular to the releases analysed from Sizewell; for different releases from different locations the sensitivity may change significantly. In the earlier study and analysis was undertaken of the impact on the predicted consequences of potential overestimates in the release fractions of radionuclides. Since the results of that study were published some relatively minor numerical errors have been identified. While none of these affects the conclusions reached in that study the opportunity has been taken in this report to present revised values for those results known to be in error. This revised text and results are presented as an appendix to this report and they replace the corresponding material in the earlier study. (author)

  6. 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.

  7. Evaluation of an accident risk of advanced pwr type reactors. Methods and results of a comprehensive probabilistic safety analysis (PSA)

    International Nuclear Information System (INIS)

    Nuclear power plant operators in Germany, and especially the opeator of GKN 2, support the GRS by providing information updates of the probabilistic methods of safety assessment. The report presents the investigations and results of probabilistic safety analyses for PWR power plants according to the current state of the art. The PSA uses methods that can be utilized by a wide range of experienced users and which have been tested by a 12-year continuous PSA for a nuclear power plant with an advanced PWR reactor. (orig.)

  8. Practice and prospect of advanced fuel management and fuel technology application in PWR in China

    International Nuclear Information System (INIS)

    Since Daya Bay nuclear power plant implemented 18-month refueling strategy in 2001, China has completed a series of innovative fuel management and fuel technology projects, including the Ling Ao Advanced Fuel Management (AFM) project (high-burnup quarter core refueling) and the Ningde 18-month refueling project with gadolinium-bearing fuel in initial core. First, this paper gives brief introduction to China's advanced fuel management and fuel technology experience. Second, it introduces practices of the advanced fuel management in China in detail, which mainly focuses on the implementation and progress of the Ningde 18-month refueling project with gadolinium-bearing fuel in initial core. Finally, the paper introduces the practices of advanced fuel technology in China and gives the outlook of the future advanced fuel management and fuel technology in this field. (author)

  9. Investigation of the ex-core noise induced by fuel assembly vibrations in the Ringhals-3 PWR

    International Nuclear Information System (INIS)

    Highlights: • The effect of cycle burnup on the ex-core noise induced by fuel assembly vibrations in a PWR was investigated. • No general monotonic variation of APSD during cycle burnup was found. • The increase of APSD occurs primarily for the vibrations of peripheral assemblies. • The ex-core noise is dominated by peripheral assemblies. • For the vibration of several assemblies distributed throughout the core in Model 2, a monotonic increase of APSD was found. - Abstract: The effect of cycle burnup on the ex-core detector noise at the frequency of the pendular core barrel vibrations in the Ringhals-3 PWR core was investigated using a neutron noise simulator. The purpose of the investigations was to confirm or disprove a hypothesis raised by Sweeney et al. (1985) that fuel assembly vibrations could affect the ex-core detector noise and cause the corresponding peak in the auto power spectral density (APSD) to increase during the cycle due to the effects of fuel burnup, the change of boron concentration, flux redistribution etc. Numerical calculations were performed by modelling the vibrations of fuel assemblies at different locations in the core and calculating the induced neutron noise at three burnup steps. The APSD of the ex-core detector noise was evaluated with the assumption of vibrations either along a straight-line or along a random two-dimensional trajectory, with two different representations of the cross section perturbations caused by the vibrations. The results show the obvious effect of in-core fuel vibrations on the ex-core detector noise, but the monotonic increase of the APSD does not occur for all fuel elements, vibration types and cross section perturbation models. Such an increase of the of APSD occurs predominantly for peripheral assemblies with one of the perturbation models. However, assuming simultaneous vibrations of a number of fuel assemblies uniformly distributed over the core with random vibrations and the more realistic

  10. 3D coarse mesh NEM embedded with 2D fine mesh NDOM for PWR core analysis - 032

    International Nuclear Information System (INIS)

    The paper describes an algorithm of 3D multi-group coarse mesh Nodal Expansion Method (NEM) embedded with 2D multi-group fine mesh Nodal Discrete Ordinate Method (NDOM). The main idea of the algorithm is to substitute the radial part of 3D NEM inner iteration, which is based on coarse mesh diffusion theory, with 2D NDOM inner iteration, which is based on transport theory. Taking advantages of both NEM and NDOM, the algorithm efficiently models the heterogeneous pin-by-pin layout in radial planes of PWR core and overcomes the challenges of computer memory and computation time, which are inherent bottlenecks of 3D fine mesh discrete ordinate method (Sn). 2 prototype codes MGNSNM and MGNEM are developed, which are based on 2D multi-group NDOM and 3D multi-group NEM respectively; the final computer code HANWIND is integrated based on the above 2 prototype codes. Numerical experiments on benchmark problem OECD/NEA-2D C5G7MOX and a self-established benchmark problem of 2-loop PWR 3D core are summarized in the paper. (authors)

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

    International Nuclear Information System (INIS)

    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

  12. IRIS-50. A 50 MWe advanced PWR design for smaller, regional grids and specialized applications

    International Nuclear Information System (INIS)

    IRIS is an advanced, medium-power (1000 MWt or ∼335 MWe) advanced PWR design of integral configuration, that has gained wide recognition due to its innovative 'safety-by-design' safety approach. In spite of its smaller size compared to large monolithic nuclear power plants, it is economically competitive due to its simplicity and advantages of modular deployment. However, the optimum power level for a class of specific applications (e.g., power generation in small regional isolated grids; water desalination and biodiesel production at remote locations; autonomous power source for special applications, etc.) may be even lower, of the order of tens rather than hundreds of MWe. The simple and robust IRIS 335 MWe design provides a solid basis for establishing a 20-100 MWe design, utilizing the same safety and economics principles, so that it will retain economic attractiveness compared to other alternatives of the same power level. A conceptual 50 MWe design, IRIS-50, was initially developed and then assessed in a 2001 report to the US Congress on small and medium reactors, as a design mature enough to have deployment potential within a decade. In the meantime, while the main efforts have focused on the 335 MWe design completion and licensing, parallel efforts have progressed toward the preliminary design of IRIS-50. This paper summarizes the main IRIS-50 features and presents an update on its design status. (author)

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

    International Nuclear Information System (INIS)

    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 241Am might play the decisive role of burnable absorber to reduce the initial reactivity excess while the short-lived nuclides 242Cm and 244Cm decay into even plutonium isotopes, thus increasing the extent of denaturation for primary fissile 239Pu 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 241Am, 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% 238Pu 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)

  14. 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)

  15. Study of advanced LWR cores for effective use of plutonium and MOX physics experiments

    International Nuclear Information System (INIS)

    Advanced technologies of full MOX cores have been studied to obtain higher Pu consumption based on the advanced light water reactors (APWRs and ABWRs). For this aim, basic core designs of high moderation lattice (H/HM ∼5) have been studied with reduced fuel diameters in fuel assemblies for APWRs and those of high moderation lattice (H/HM ∼6) with addition of extra water rods in fuel assemblies for ABWRs. The analysis of equilibrium cores shows that nuclear and thermal hydraulic parameters satisfy the design criteria and the Pu consumption rate increases about 20 %. An experimental program has been carried out to obtain the core parameters of high moderation MOX cores in the EOLE critical facility at the Cadarache Centre as a joint study of NUPEC, CEA and CEA's industrial partners. The experiments include a uranium homogeneous core, two MOX homogeneous cores of different moderation and a PWR assembly mock up core of MOX fuel with high moderation. The program was started from 1996 and will be completed in 2000. (author)

  16. Analysis of core-melt states for the development of detection methods for filling level change and deformation of the core in PWR-type reactors

    International Nuclear Information System (INIS)

    The project ''noninvasive status monitoring of nuclear reactors for detection of filling level changes and core deformation'' (NIZUK) is aimed to develop a measuring system for the core status diagnosis during severe accidents in PWR-type reactors. For the development of an appropriate measuring technology the knowledge on the processes during the in-vessel phase of the accident sequence is of main importance. Using the analysis of the accident sequence nine in-vessel phases were defined that are the basis for the development of the measuring system. The differences between the individual core-melt states include the different core geometries and a varying gamma radiation distribution at the reactor pressure vessel outer surface. Especially the appearance of local flow-off paths during a late in-vessel phase requires that several measuring probes with gamma radiation sensors have to be installed around the reactor pressure vessel in order to detect the gamma radiation distribution at the outside. The definition of further core-melt states would be possible in case of a re-flooding of the reactor pressure vessel. However, the increasing filling level would not significantly change the core deformation and the gamma distribution at the outside.

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

    Energy Technology Data Exchange (ETDEWEB)

    Höhne, Thomas, E-mail: T.Hoehne@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Institute of Fluid Dynamics, P.O. Box 510119, D-01328 Dresden (Germany); Grahn, Alexander; Kliem, Sören; Rohde, Ulrich; Weiss, Frank-Peter [Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Institute of Fluid Dynamics, P.O. Box 510119, D-01328 Dresden (Germany)

    2013-05-15

    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

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

    International Nuclear Information System (INIS)

    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

  19. A quantitative comparison of loading pattern optimization methods for in-core fuel management of PWR

    International Nuclear Information System (INIS)

    The performance of several loading pattern(LP) optimization methods was quantitatively compared through a benchmark problem of PWR LP optimization. The simulated annealing(SA) method, the genetic algorithms(GA) method, the direct search(DS) method based on assembly multiple shuffling and the binary exchange(BE) method based on fuel assembly binary exchange were investigated as candidates for the optimization techniques. Hybrid strategy which combined different optimization methods was newly proposed, and the performances of two different new hybrid methods, which combined DS with BE and GA with BE were examined. From the results of the LP optimization benchmark problem, the superiority and inferiority of each method were clarified. Furthermore, it was demonstrated that the GA+BE hybrid strategy performed best among these methods. By combining GA with BE, the weaknesses of these two methods were compensated for with each other and the optimization performance was improved significantly. Therefore, the GA+BE hybrid method is quite effective for the LP optimization problems of PWR. (author)

  20. Coupled neutronic thermo-hydraulic analysis of full PWR core with Monte-Carlo based BGCore system

    International Nuclear Information System (INIS)

    Highlights: → New thermal-hydraulic (TH) feedback module was integrated into the MCNP based depletion system BGCore. → A coupled neutronic-TH analysis of a full PWR core was performed with the upgraded BGCore system. → The BGCore results were verified against those of 3D nodal diffusion code DYN3D. → Very good agreement in major core operational parameters between the BGCore and DYN3D results was observed. - Abstract: BGCore reactor analysis system was recently developed at Ben-Gurion University for calculating in-core fuel composition and spent fuel emissions following discharge. It couples the Monte Carlo transport code MCNP with an independently developed burnup and decay module SARAF. Most of the existing MCNP based depletion codes (e.g. MOCUP, Monteburns, MCODE) tally directly the one-group fluxes and reaction rates in order to prepare one-group cross sections necessary for the fuel depletion analysis. BGCore, on the other hand, uses a multi-group (MG) approach for generation of one group cross-sections. This coupling approach significantly reduces the code execution time without compromising the accuracy of the results. Substantial reduction in the BGCore code execution time allows consideration of problems with much higher degree of complexity, such as introduction of thermal hydraulic (TH) feedback into the calculation scheme. Recently, a simplified TH feedback module, THERMO, was developed and integrated into the BGCore system. To demonstrate the capabilities of the upgraded BGCore system, a coupled neutronic TH analysis of a full PWR core was performed. The BGCore results were compared with those of the state of the art 3D deterministic nodal diffusion code DYN3D. Very good agreement in major core operational parameters including k-eff eigenvalue, axial and radial power profiles, and temperature distributions between the BGCore and DYN3D results was observed. This agreement confirms the consistency of the implementation of the TH feedback module

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  3. Analysis of a rod withdrawal in a PWR core with the neutronic- thermalhydraulic coupled code RELAP/PARCS and RELAP/VALKIN

    International Nuclear Information System (INIS)

    The Reactor Ejection Accident (REA) belongs to the Reactor Initiated Accidents (RIA) category of accidents and it is part of the licensing basis accident analyses required for pressure water reactors (PWR). The REA at hot zero power (HZP) is characterized by a single rod ejection from a core position with a very low power level. The evolution consists basically of a continuous reactivity insertion. The main feature limiting the consequences of the accident in a PWR is the Doppler Effect. To check the performance of the coupled code RELAP5/PARCS2.5 and RELAP5/VALKIN a REA in Trillo NPP is simulated. These analyses will allow knowing more accurately the PWR real plant phenomenology in the RIA most limiting conditions. (authors)

  4. Improving PWR core simulations by Monte Carlo uncertainty analysis and Bayesian inference

    CERN Document Server

    Castro, Emilio; Buss, Oliver; Garcia-Herranz, Nuria; Hoefer, Axel; Porsch, Dieter

    2016-01-01

    A Monte Carlo-based Bayesian inference model is applied to the prediction of reactor operation parameters of a PWR nuclear power plant. In this non-perturbative framework, high-dimensional covariance information describing the uncertainty of microscopic nuclear data is combined with measured reactor operation data in order to provide statistically sound, well founded uncertainty estimates of integral parameters, such as the boron letdown curve and the burnup-dependent reactor power distribution. The performance of this methodology is assessed in a blind test approach, where we use measurements of a given reactor cycle to improve the prediction of the subsequent cycle. As it turns out, the resulting improvement of the prediction quality is impressive. In particular, the prediction uncertainty of the boron letdown curve, which is of utmost importance for the planning of the reactor cycle length, can be reduced by one order of magnitude by including the boron concentration measurement information of the previous...

  5. Analytical Transmission Electron Microscopy Characterization of Stress Corrosion Cracks in an Irradiated Type 316 Stainless Steel Core Component of PWR

    International Nuclear Information System (INIS)

    Irradiation-assisted stress-corrosion cracking (IASCC) of a cold-worked type 316 stainless steel baffle/former bolt from a pressurized-water reactor (PWR) was investigated by analytical transmission electron microscopy (ATEM). Nanometer-resolution methods for feature-specific analysis were used to characterize irradiation- and corrosion-affected microstructures of the crack tip. The work is part of an international cooperative program to characterize light-water-reactor core components that experience IASCC. This is the first detailed ATEM examination of in-service cracks in neutron-irradiated austenitic stainless steel. The bolt cracked during 20-year service at an estimated temperature of 299 C and 8.5 dpa dose. Cross-section samples prepared by dimple-grinding and ion micro-milling were examined in a 200 kV field-emission-gun TEM. A wide variety of high-resolution ATEM methods for microstructural observation, sub-nm electron-probe analysis and diffraction were used to characterize the cracks and adjacent alloy matrix. The alloy appeared entirely austenitic with primary grain sizes from ∼ 50 to 100 μm and contained deformation twins from cold working. Coarse stringer inclusions of MnS and complex oxides with particle sizes to 10's of μm were observed to intersect the crack. High densities of remnant cold-work dislocations, 1- to 30-nm radiation-induced Frank loops and also sub-nm gas bubbles were observed throughout the alloy matrix. Other radiation-induced defects in the grain interiors were scattered 2-5 nm voids (negligible total swelling) and 2- to 4-nm γ' particles. Grain boundaries in the alloy were precipitate-free, but had ∼ 4-nm-wide zones of high Ni and Si enrichment and corresponding Mo, Cr and Fe depletions due to radiation-induced segregation. The cracking occurred along high-angle grain boundaries (excepting coherent Σ3-type twin boundaries). Cracks were filled with oxide corrosion products consisting of fine-grained, Cr-rich spinel along

  6. CFD simulation of fibre material transport in a PWR core under loss of coolant conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, Thomas; Grahn, Alexander; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany); Weiss, Frank-Peter [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Garching (Germany)

    2011-03-15

    During a postulated cold leg LOCA with hot leg ECC injection, a limited amount of small fractions of the insulation material after passing the sump strainers can 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. The CFD simulations show that after starting the sump mode, the ECC water injected through the hot legs flows down into the core at so-called 'brake through channels' located at the outer core region where the downward leg of the convection role had established. The hotter, lighter coolant rises in the center of the core. As a consequence, the insulation material is preferably deposited at the uppermost spacer grids positioned in the break through zones. This means that at the beginning the fibers are not uniformly deposited over the core cross section. (orig.)

  7. Recalculation of simulated post-scram core power decay curve for use in ROSA-IV/LSTF experiments on PWR small-break LOCAs and transients

    International Nuclear Information System (INIS)

    Simulated post-scram core power decay curve for use in Large Scale Test Facility (LSTF) tests has been calculated on a best-estimate basis, particularly in two points, i.e. estimation of the delayed neutron fission power and consideration of the stored heat in a pressurized water reactor (PWR) fuel rod. The New Power Curve provides a LSTF heater rod with the heat transfer rate from a PWR fuel rod that was estimated for a typical pressure transient during a PWR small-break loss of coolant accident. This approach neglects conservatively the effect of stored heat release from the LSTF heater rod considering that there is large uncertainty in the thermal conductivity of outer insulator in the LSTF heater rod. When the New Power Curve is used as the LSTF core power curve, the heat transfer rate from a LSTF heater rod gives a little conservative values as compared with the heat transfer rate from a PWR fuel rod. (author)

  8. Verification of the Monte Carlo code RMC with a whole PWR MOX/UO2 core benchmark

    International Nuclear Information System (INIS)

    Several types of V and V work are being carried out for the Reactor Monte Carlo code RMC, including the heterogeneous whole core configurations. In this paper, a whole PWR MOX/UO2 core benchmark which contains both UO2 and MOX assemblies with different enrichments and various burn-up points is chosen to verify RMC's criticality calculation capability, and the results of RMC and other codes are discussed and compared, such as eigenvalues, assembly power distributions, pin power distributions and so on. The discrepancies in eigenvalues and power distributions are satisfactory, which proves the accuracy of RMC's criticality calculation. Also, the influences of different cross-section libraries are discussed upon the results of RMC. Besides these results, the detailed comparisons between RMC and MCNP with the same ENDF/B-VII.0 cross-section library are carried out in this paper, including the comparisons of control rod worths calculated by both RMC and MCNP. According to the results, RMC and MCNP agree quite well in eigenvalues, power distributions and other results. The discrepancies of eigenvalues and control rod worth are fairly small and the relative differences of assembly and pin power distributions are acceptable. All these results contribute to the conclusion that the criticality calculation performance of RMC is accurate and excellent. (author)

  9. A preliminary study of thorium and transuranic advanced fuel cycle utilization in PWR

    International Nuclear Information System (INIS)

    A typical PWR fuel element considering (TRU-Th) cycle was simulated. The study analyzed the behaviour of the thorium insertion spiked with reprocessed fuel considering different enrichments that varied from 5.5% to 7.0%. The reprocessed fuels were obtained using the ORIGEN 2.1 code from a burned PWR standard fuel (33,000 MWd/tHM burned), with 3.1% of initial enrichment, which was remained in the cooling pool for five years. The Kerf, hardening spectrum, and the fuel evolution during the burnup were evaluated. This study was performed using the SCALE 6.0. (author)

  10. Quasi-static core liquid level depression and long-term core uncovery during a PWR LOCA

    International Nuclear Information System (INIS)

    The possibility exists that the core liquid level can be depressed quasi-statically during long-term plant cooldown following a cold-leg-break loss-of-coolant accident in a pressurized-water reactor. Such a level depression can take place if the core steaming rate is almost balanced with the steam condensation rate in the reactor coolant system (RCS). Integral experiments simulating three steam condensation modes in the RCS have shown that quasi-static level depressions can happen and thus have the potential for causing a prolonged uncovery and heatup of the core upper regions

  11. Core liquid level depression due to manometric effect during PWR small break LOCA

    International Nuclear Information System (INIS)

    The 10, 5 and 2.5 % cold leg break loss-of-coolant accident experiments were conducted by using the Large Scale Test Facility (LSTF) of the Rig of Safety Assessment (ROSA)-IV program. In the early stage of the 5 % break experiment, the core collapsed liquid level was depressed nearly to the core bottom and the dryout of the core was observed. However, the core liquid level depression without the core dryout was observed in the 10 and 2.5 % break experiments. In the three break experiments, the core liquid levels were recovered just after the loop seal clearing. The manometric effect due to the liquid seal formation in the loop seal and the liquid holdup in the steam generator (SG) U-tubes upflow-side caused a depression of the core collapsed liquid level. The liquid holdup in the U-tubes upflow-side was observed after the termination of the two-phase circulation due to the phase separation at the U-tubes top. The counter current flow limiting (CCFL) and the condensation of steam was considered to be the main reason for the liquid holdup. In the 10, 5 and 2.5 % break experiments, the termination of the two-phase circulation and the loop seal clearing were observed approximately at 40 ∼ 60 % and 30 % mass inventory in the primary system, respectively. (author)

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

    International Nuclear Information System (INIS)

    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)

  13. Out-pile test of non-instrumented capsule for the advanced PWR fuel pellets in HANARO irradiation test

    International Nuclear Information System (INIS)

    Non-instrumental capsule were designed and fabricated to irradiate the advanced pellet developed for the high burn-up LWR fuel in the HANARO in-pile capsule. This capsule was out-pie tested at Cold Test Loop-I in KAERI. From the pressure drop test results, it is noted that the flow velocity across the non-instrumented capsule of advanced PWR fuel pellet corresponding to the pressure drop of 200 kPa is measured to be about 7.45 kg/sec. Vibration frequency for the capsule ranges from 13.0 to 32.3 Hz. RMS displacement for non-instrumented capsule of advanced PWR fuel pellet is less than 11.6 μm, and the maximum displacement is less that 30.5 μm. The flow rate for endurance test were 8.19 kg/s, which was 110% of 7.45 kg/s. And the endurance test was carried out for 100 days and 17 hours. The test results found not to the wear satisfied to the limits of pressure drop, flow rate, vibration and wear in the non-instrumented capsule

  14. 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)

  15. Effect of the fluid structure interaction on the dynamic response of a PWR reactor core

    International Nuclear Information System (INIS)

    The criterion retained for the design of the fuel assembly mixing grids under accidental dynamic loads (earthquake and LOCA) consists of checking that the maximum impact force on the grids does not exceed their resistance to dynamic buckling. The first stage in design is therefore the calculation of the reaction of the core to these accidental stresses. This article describes the approach which determines lateral loading on the fuel assembly mixing grids during earthquake. In this formulation, forces induced by the fluid structure interaction are taking into account. And a comparison between the dynamic response of the core with or without taking into account the fluid structure interaction is presented

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

    International Nuclear Information System (INIS)

    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)

  17. Optimization of enthalpy transport coefficient for a simplified PWR core thermal-hydraulic calculation

    International Nuclear Information System (INIS)

    A simplified thermal-hydraulic program was developed to quickly calculate the minimum Departure from a Nucleate Boiling Ratio (DNBR) in a pressurized water reactor for use in online core monitoring and protection systems. The simplified thermal-hydraulic program calculates the minimum DNBR based on a lumped four-channel core model. The conservation equations for the four channels were derived using three-dimensional transport coefficients which are used to modify the calculation of the radial transport of the enthalpy and momentum. There are three transport coefficients for the enthalpy, the axial velocity and the pressure. A previous detailed subchannel analysis showed that the enthalpy transport coefficient is strongly dependent on the core operating conditions. This study examines the variation of the enthalpy transport coefficient for two fuel types in wide range of the core operating conditions for a Korean optimized power reactor, OPR 1000. The average enthalpy transport coefficient for the no-vane fuel shows the peak at approximately axial shape index (Asi) of +0.4 but for the mixing-vane fuel appears to gradually decrease as the Asi increases. The enthalpy transport coefficient of greater than 50.0 was found to minimize the minimum DNBR error for the fast minimum DNBR calculation. (Author)

  18. Bayesian inference along Markov Chain Monte Carlo approach for PWR core loading pattern optimization

    International Nuclear Information System (INIS)

    Highlights: ► The BIMCMC method performs very well and is comparable to GA and PSO techniques. ► The potential of the technique is very well for optimization. ► It is observed that the performance of the method is quite adequate. ► The BIMCMC is very easy to implement. -- Abstract: Despite remarkable progress in optimization procedures, inherent complexities in nuclear reactor structure and strong interdependence among the fundamental indices namely, economic, neutronic, thermo-hydraulic and environmental effects make it necessary to evaluate the most efficient arrangement of a reactor core. In this paper a reactor core reloading technique based on Bayesian inference along Markov Chain Monte Carlo, BIMCMC, is addressed in the context of obtaining an optimal configuration of fuel assemblies in reactor cores. The Markov Chain Monte Carlo with Metropolis–Hastings algorithm has been applied for sampling variable and its acceptance. The proposed algorithm can be used for in-core fuel management optimization problems in pressurized water reactors. Considerable work has been expended for loading pattern optimization, but no preferred approach has yet emerged. To evaluate the proposed technique, increasing the effective multiplication factor Keff of a WWER-1000 core along flattening power with keeping power peaking factor below a specific limit as a first test case and flattening of power as a second test case are considered as objective functions; although other variables such as burn up and cycle length can also be taken into account. The results, convergence rate and reliability of the new method are compared to published data resulting from particle swarm optimization and genetic algorithm; the outcome is quite promising and demonstrating the potential of the technique very well for optimization applications in the nuclear engineering field.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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. BEACON TSM application system to the operation of PWR reactors

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

  3. Exploratory study of molten core material/concrete interactions, July 1975--March 1977. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Powers, D.A.; Dahlgren, D.A.; Muir, J.F.; Murfin, W.D.

    1978-02-01

    An experimental study of the interaction between high-temperature molten materials and structural concrete is described. The experimental efforts focused on the interaction of melts of reactor core materials weighing 12 to 200 kg at temperatures 1700 to 2800/sup 0/C with calcareous and basaltic concrete representative of that found in existing light-water nuclear reactors. Observations concerning the rate and mode of melt penetration into concrete, the nature and generation rate of gases liberated during the interaction, and heat transfer from the melt to the concrete are described. Concrete erosion is shown to be primarily a melting process with little contribution from mechanical spallation. Water and carbon dioxide thermally released from the concrete are extensively reduced to hydrogen and carbon monoxide. Heat transfer from the melt to the concrete is shown to be dependent on gas generation rate and crucible geometry. Interpretation of results from the interaction experiments is supported by separate studies of the thermal decomposition of concretes, response of bulk concrete to intense heat fluxes (28 to 280 W/cm/sup 2/), and heat transfer from molten materials to decomposing solids. The experimental results are compared to assumptions made in previous analytic studies of core meltdown accidents in light-water nuclear reactors. A preliminary computer code, INTER, which models and extrapolates results of the experimental program is described. The code allows estimation of the effect of physical parameters on the nature of the melt/concrete interaction.

  4. Introduction of thorium in under-moderated PWR cores and optimization to enhance the conversion factor

    International Nuclear Information System (INIS)

    Studies have been undertaken in order to optimize the use of natural uranium in Pressurized Water Reactors. Thus, a core with high conversion factor (CF) and hexagonal lattice has been developed. The CF is defined as the ratio between the fissile isotopes production rate and their consumption rate. To reach high CF, the Moderation Ratio (MR) has to be drastically reduced. An optimum between thermal-hydraulic limitations and neutronic performances is a MRcell of 0.83. A core with depleted uranium and plutonium shows promising performances especially thanks to both axial and radial heterogeneities. In addition to that, 232Th is a fertile isotope similar to 238U which produces 233U by capture. This fissile isotope has two main advantageous properties. First it is a stable element (in comparison to 241Pu which decays into 241Am with a period of about 15 years); secondly, it has a neutron yield per neutron absorbed higher than 239Pu in thermal and epithermal spectrum. The use of thorium in substitution of depleted uranium is interesting from the CF point of view and has not yet been studied in France in this kind of concept. (author)

  5. 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)

    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)

  6. Advanced water processing system (AWPS), including advanced filtration system (AFS) and advanced ion selective system (AISS) for improved utility (PWR/BWR) water processing performance

    International Nuclear Information System (INIS)

    The advanced water processing system (AWPS) has the potential for wide spread success on a worldwide scale in both PWR and BWRs. The AWPS incorporates the advanced features (patent pending) of advanced filtration and advanced ion selective technologies (patented). Typical problems encountered in current filtration systems include: (1) poor effluent quality, (2) short run lengths on filters, (3) frequent filter change-outs/backwashes, (4) large waste volumes, and (5) failed filter cartridges. The advanced filtration system (AFS) features reduced waste production per million gallons of water processed, cleaner water for recycle or release to the environment, filter element volume 100 times less than that of competitive filters, and a far lower capital cost compared to systems with similar performance. The AWPS should be of interest to plants that are upgrading, or to new plants to lower both their capital and operating costs, as well as total curie discharge levels. In addition, the AWPS will function in non-nuclear, as well as nuclear, applications of water purification, specially where pre coat filtration/ion exchange or reverse osmosis (RO) is being applied to process water with high concentrations of colloidal contaminants. Pilot testing has been successfully completed in the U. S. at the Byron (PWR), LaSalle (BWR), and Dresden(BWR) nuclear plants for Commonwealth Edison, and the Bruce several spent filters in a High Integrated Container these bench- and pilot-scale demonstrations will be presented herein. Full-scale designs or systems have been shipped to these locations. In all cases, the testing demonstrated: (1) longer run lengths (300,000 gallons between backwashes--a 100 fold improvement), (2) recoverability of cartridge filters after backwash (cartridge lives of approximately 6 months to a year--a 5 to 10 fold improvement in filter life), (3) large removal efficiencies for colloidal particles (reduced discharge curies), and (4) reduced waste volumes

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

    International Nuclear Information System (INIS)

    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)

  8. PWR rod ejection accident: uncertainty analysis on a high burn-up core configuration

    Energy Technology Data Exchange (ETDEWEB)

    Le Pallec, J.C.; Studer, E.; Royer, E. [CEA Saclay, Direction de l' Energie Nucleaire, Service d' Etudes de Reacteurs et de Modelisation Avancee (DEN/SERMA), 91 - Gif sur Yvette (France)

    2003-07-01

    With the increasing of the discharge burn-up assembly, the rod ejection accident (REA) methodology based on the analyse of the hot spot from a decoupling methods of calculation does not allow to ensure the respect of safety criteria. The main reason is that the irradiated fuel certainly less solicited thermally is in the other hand more sensitive to a transient due to a rod ejection. Thus, the hot spot is not necessarily the sensitive point of the core. In the framework of high burn-up configurations, a new methodology tends to replace the former. It characterizes by the use of a best-estimate 3-dimensional modelling: coupling of the thermal hydraulics and neutronics, taking in account fuel properties depending on irradiation. To ensure the conservatism of the modelling response, this new approach has to be followed by an uncertainties analysis. Inputs from the benchmark RIA TMI-1 conducted by IRSN (France), NRC (United State of America) and KI (Russian) are used to perform a first analysis. The response of the modelling is the enthalpy deposited in an assembly. The analysis is based on the Design of Experiments (DoE) that permits to measure the weight of the main parameters and their interactions on the response. These last cannot be disregarded because they represent up to 20% of the penalizing uncertainty. This study shows that the main fuel modifications due to irradiation (radial power distribution, thermal properties degradation) have to be taken into account in a realistic thermal modelling during a strong transient.

  9. Neutronic Analysis on the Fuel Pin Reshuffling Options for PWR In-Core Fuel Management

    International Nuclear Information System (INIS)

    Such estimation has been verified by comparing with the neutronic performance of the reference design. Two feasible scenarios have been studied by targeting on the improvement of the uniform flux spatial distribution and on the enhancement of neutron economy leading to economic incentives. In the first scenario, for making flux more uniform spatial distribution, the existing fuel pins are relocated to have fuel pins burnt more evenly. It is expected to result in minimization of nuclear power peaking while maximizing neutron economy at the core edge. Secondly, with the help of the intra-fuel assembly reshuffling option, the spent fuel pins could be reused again by combining other available twice burnt fuel pins so that numbers of new fresh FAs as well as discharged FAs will be reduced. In scenario-1, the operating time was merely somewhat increased for few minutes by keeping enough safety margins. The secnario-2 was proved to reduce 4 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. In brief, the OPR-S1-a and the OPR-S1-b designs could not bring distinctive economic benefits if only few fuel assemblies would be treated through the intra-fuel assembly reshuffling strategy. The OPR-S2 design was proved to be acceptable except slightly larger nuclear power peaking factor. Because of reduction fresh fuel assembly loading, that will bring more economic incentives. It is, though, still left to investigate more into the intra-fuel assembly reshuffling despite hardship for handling massive numbers of fuel pins, and their radiotoxicity and complexity. If these obstacles could be overcome, it would be worth to examine significant economy incentives and safety performance benefits

  10. Calculations concerning the capability of passive recombiners to control hydrogen concentration in the containment of an advanced PWR

    International Nuclear Information System (INIS)

    The Department of Mechanical and Nuclear Constructions of the University of Pisa has developed a computer code, HOCRA, which is able to make an initial evaluation of the capability of catalytic recombiners to remove hydrogen from the atmosphere of the safety containments of nuclear reactors in accident conditions. The code allows the analysis of the average concentration transient of hydrogen in a generic compartment of a safety containment in a nuclear reactor. The software is structured into two groups. The first, mode-1, analyses the average concentration in all the free volume of the containment before a possible venting., whereas the second, mode-2, analyses the average concentration transient in a containment compartment, assuming input and output flow rates into and from the compartment itself The first part of this paper outlines the physical and mathematical model of the code, the second part reports calculations made for an advanced PWR in cooperation with ENEL. (author)

  11. Performance of Core Exit Thermocouple for PWR Accident Management Action in Vessel Top Break LOCA Simulation Experiment at OECD/NEA ROSA Project

    Science.gov (United States)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

    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 the 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 reasons 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.

  12. CAE advanced reactor demonstrators for CANDU, PWR and BWR nuclear power plants

    International Nuclear Information System (INIS)

    CAE, a private Canadian company specializing in full scope flight, industrial, and nuclear plant simulators, will provide a license to IAEA for a suite of nuclear power plant demonstrators. This suite will consist of CANDU, PWR and BWR demonstrators, and will operate on a 486 or higher level PC. The suite of demonstrators will be provided to IAEA at no cost to IAEA. The IAEA has agreed to make the CAE suite of nuclear power plant demonstrators available to all member states at no charge under a sub-license agreement, and to sponsor training courses that will provide basic training on the reactor types covered, and on the operation of the demonstrator suite, to all those who obtain the demonstrator suite. The suite of demonstrators will be available to the IAEA by March 1997. (author)

  13. Development of inherent core technologies for advanced reactor

    International Nuclear Information System (INIS)

    Recently, the developed countries made their effort on developing the advanced reactor which will result in significantly enhanced safety and economy. However, they will protect the advanced reactor and its design technology with patent and proprietary right. Therefore, it is very important to develop our own key core concepts and inherent core design technologies which can form a foundation of indigenous technologies for development of the domestic advanced reactor in order to keep the superiority in the nuclear plant building market among the developing countries. In order to provide the basic technology for the core design of advanced reactor, this project is for developing the inherent core design concepts with enhanced safety and economy, and associated methodologies and technologies for core analyses. The feasibility study of constructing domestic critical facilities are performed by surveying the status and utilization of foreign facilities and by investigating the demand for domestic facilities. The research results developed in this project, such as core analysis methodologies for hexagonal core, conceptual core design based on hexagonal fuel assemblies and soluble boron core design and control strategies, will provide a technical foundation in developing core design of domestic advanced reactor. Furthermore, they will strengthen the competitiveness of Korean nuclear technology. We also expect that some of the design concepts developed in this project to improve the reactor safety and economy can be applicable to the design of advanced reactor. This will significantly reduce the public anxiety on the nuclear power plant, and will contribute to the economy of construction and operation for the future domestic reactors. Even though the critical facility will not be constructed right now, the investigation of the status and utilization of foreign critical facility will contribute to the future critical facility construction. (author). 150 refs., 34 tabs., 103

  14. Development of inherent core technologies for advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Keung Koo; Noh, J.M.; Hwang, D.H. [and others

    1999-03-01

    Recently, the developed countries made their effort on developing the advanced reactor which will result in significantly enhanced safety and economy. However, they will protect the advanced reactor and its design technology with patent and proprietary right. Therefore, it is very important to develop our own key core concepts and inherent core design technologies which can form a foundation of indigenous technologies for development of the domestic advanced reactor in order to keep the superiority in the nuclear plant building market among the developing countries. In order to provide the basic technology for the core design of advanced reactor, this project is for developing the inherent core design concepts with enhanced safety and economy, and associated methodologies and technologies for core analyses. The feasibility study of constructing domestic critical facilities are performed by surveying the status and utilization of foreign facilities and by investigating the demand for domestic facilities. The research results developed in this project, such as core analysis methodologies for hexagonal core, conceptual core design based on hexagonal fuel assemblies and soluble boron core design and control strategies, will provide a technical foundation in developing core design of domestic advanced reactor. Furthermore, they will strengthen the competitiveness of Korean nuclear technology. We also expect that some of the design concepts developed in this project to improve the reactor safety and economy can be applicable to the design of advanced reactor. This will significantly reduce the public anxiety on the nuclear power plant, and will contribute to the economy of construction and operation for the future domestic reactors. Even though the critical facility will not be constructed right now, the investigation of the status and utilization of foreign critical facility will contribute to the future critical facility construction. (author). 150 refs., 34 tabs., 103

  15. 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)

    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

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

    International Nuclear Information System (INIS)

    The computational capability of efficiently and accurately evaluate reactor core attributes (i.e., keff 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.)

  17. Advanced numerical techniques in core simulations

    International Nuclear Information System (INIS)

    The whole core simulations are one of the most CPU intensive calculations in reactor physics design and analyses. For a designer it is imperative to perform these calculations with good accuracy and in least time possible to try out various options. It is important for the code developers to use techniques involving minimum approximations and to use most recent numerical methods applied in tandem with huge computing power available today. In the presented paper, some of these methods are discussed. (author)

  18. 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.

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

    International Nuclear Information System (INIS)

    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.

  20. Design and manufacturing of non-instrumented capsule for advanced PWR fuel pellet irradiation test in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. H.; Lee, C. B.; Song, K. W. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    This project is preparing to irradiation test of the developed large grain UO{sub 2} fuel pellet in HANARO for pursuit fuel safety and high burn-up in 'Advanced LWR Fuel Technology Development Project' as a part Nuclear Mid and Long-term R and D Program. On the basis test rod is performed the nuclei property and preliminary fuel performance analysis, test rod and non-instrumented capsule are designed and manufactured for irradiation test in HANARO. This non-instrumented irradiation capsule of Advanced PWR Fuel pellet was referred the non-instrumented capsule for an irradiation test of simulated DUPIC fuel in HANARO(DUPIC Rig-001) and 18-element HANARO fuel, was designed to ensure the integrity and the endurance of non-instrumented capsule during the long term(2.5 years) irradiation. To irradiate the UO{sub 2} pellets up to the burn-up 70 MWD/kgU, need the time about 60 months and ensure the integrity of non-instrumented capsule for 30 months until replace the new capsule. This non-instrumented irradiation capsule will be based to develope the non-instrumented capsule for the more long term irradiation in HANARO. 22 refs., 13 figs., 5 tabs. (Author)

  1. Study on transient hydrogen behavior and effect on passive containment cooling system of the advanced PWR

    International Nuclear Information System (INIS)

    A certain amount of hydrogen will be generated due to zirconium-steam reaction or molten corium concrete interaction during severe accidents in the pressurized water reactor (PWR). The generated hydrogen releases into the containment, and the formed flammable mixture might cause deflagration or detonation to produce high thermal and pressure loads on the containment, which may threaten the integrity of the containment. The non-condensable hydrogen in containment may also reduce the steam condensation on the containment surface to affect the performance of the passive containment cooling system (PCCS). To study the transient hydrogen behavior in containment with the PCCS performance during the accidents is significant for the further study on the PCCS design and the hydrogen risk mitigation. In this paper, a new developed PCCS analysis code with self-reliance intellectual property rights, which had been validated by comparison on the transients in the containment during the design basis accidents with other developed PCCS analysis code, is brief introduced and used for the transient simulation in the containment under a postulated small break LOCA of cold-leg. The results show that the hydrogen will flow upwards with the coolant released from the break and spread in the containment by convection and diffusion, and it results in the increase of the pressure in the containment due to reducing the heat removal capacity of the PCCS. (author)

  2. 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)

    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

  3. Evaluation of DNBR calculation methods for advanced digital core protection system

    International Nuclear Information System (INIS)

    This study evaluated the on-line DNBR calculation methods for an advanced digital core protection system in PWR, i.e., subchannel analysis and group-channel analysis. The subchannel code MATRA and the four-channel codes CETOP-D and CETOP2 were used here. CETOP2 is most simplified DNBR analysis code which is implemented in core protection calculator in Korea standard nuclear power plants. The detailed subchannel code TORC was used as a reference calculation of DNBR. The DNBR uncertainty and margin were compared using allowable operating conditions at Yonggwang nuclear units 3-4. The MATRA code using a nine lumping-channel model resulted in smaller mean and larger standard deviation of the DNBR error distribution. CETOP-D and CETOP2 showed conservatively biased mean and relatively smaller standard deviation of the DNBR error distribution. MATRA and CETOP-D w.r.t CETOP2 showed significant increase of the DNBR available margin at normal operating condition. Taking account for the DNBR uncertainty, MATRA and CETOP-D over CETOP2 were estimated to increase the DNBR net margin by 2.5%-9.8% and 2.5%-3.3%, respectively

  4. Plutonium recycling in PWR

    International Nuclear Information System (INIS)

    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)

  5. Non-conformity to the new regulatory requirements of measures against molten core-concrete interaction and hydrogen explosion of PWR type reactor

    International Nuclear Information System (INIS)

    On September 10, 2014, the NRA granted permission to make changes to the reactor installation of Sendai NPS units 1 and 2 so as to meet the new regulatory requirements. The author investigated review documents and others and clarified non-conformity of severe accident measures against assumed accidents (large break LOCA+ECCS water injection failure+containment spray water injection failure) of PWR type reactor to the new regulatory requirements; (1) measures for cooling molten core fallen to the bottom of the containment vessel to mitigate molten core-concrete interaction were not effective taking account of uncertainty of starting time of operator's manual operation of alternative containment spray system within 30 minutes later after meltdown and (2) measures against hydrogen explosions inside the containment vessel could not meet hydrogen concentration was less than 13% of hydrogen explosion limit taking account of uncertainty of analysis of molten core-concrete interaction. Cross check analysis and also sensitivity analysis of containment spray water injection starting delay time had been conducted, which showed analysis results of computer codes (MAAP and MELCOR) showed different characteristics and assumed delay time of 10 or 35 minutes were not consistent. Hydrogen concentration inside the containment vessel calculated with assuming 100% of Zirconium reacted with water as molten core-concrete interaction amounted to 12.6% maximum, which suggested hydrogen concentration might be larger than 13% of hydrogen explosion limit taking account of reaction of other metals such as iron of molten core. (T. Tanaka)

  6. Advanced reactor physics methods for heterogeneous reactor cores

    Science.gov (United States)

    Thompson, Steven A.

    To maintain the economic viability of nuclear power the industry has begun to emphasize maximizing the efficiency and output of existing nuclear power plants by using longer fuel cycles, stretch power uprates, shorter outage lengths, mixed-oxide (MOX) fuel and more aggressive operating strategies. In order to accommodate these changes, while still satisfying the peaking factor and power envelope requirements necessary to maintain safe operation, more complexity in commercial core designs have been implemented, such as an increase in the number of sub-batches and an increase in the use of both discrete and integral burnable poisons. A consequence of the increased complexity of core designs, as well as the use of MOX fuel, is an increase in the neutronic heterogeneity of the core. Such heterogeneous cores introduce challenges for the current methods that are used for reactor analysis. New methods must be developed to address these deficiencies while still maintaining the computational efficiency of existing reactor analysis methods. In this thesis, advanced core design methodologies are developed to be able to adequately analyze the highly heterogeneous core designs which are currently in use in commercial power reactors. These methodological improvements are being pursued with the goal of not sacrificing the computational efficiency which core designers require. More specifically, the PSU nodal code NEM is being updated to include an SP3 solution option, an advanced transverse leakage option, and a semi-analytical NEM solution option.

  7. The core design of the advanced power reactor plus (APR+)

    International Nuclear Information System (INIS)

    Advance Power Reactor Plus (APR+), a pressurized water reactor and an improved nuclear power reactor based on the Advanced Power Reactor 1400 MWe (APR1400) in Korea, has been developed with 18-month cycle operation strategy from its initial core. The APR+ core power is 4290 MWth which corresponds to a 1500 MWe class nuclear power plant. The reactor core consists of 257 fuel assemblies. Comparing with APR1400 core design, 16 fuel assemblies are added. Its cycle length is expected about 450 EFPD directly from initial core, although most of previous other plants had been started according to their annual or 15-month cycle operation schedule at their initial core and gone to 18-month after third - fourth cycle. In order to reduce the peaking power, fuel pin configurations of the assembly, are optimized by using some low enriched fuel pins and gadolinia bearings. APR+ core has been met the requirements as well as the above cycle length requirement; 1) peaking factor, 2) Negative MTC(Moderator Temperature Coefficient), 3) sufficient shutdown margin, 4) convergent Xenon stability Index. The maximum rod burnup and the discharge fuel assembly burnup are also satisfied those of the limit. It is expected to acquire the standard design approval by the end of 2012 by the Korean nuclear regulatory. (authors)

  8. Application of mini-core of a PWR within the framework benchmark of the OECD Uncertainty Analysis in Modelling

    International Nuclear Information System (INIS)

    This work studies the effect that produces the homogenization and condensation of the effective sections on jointly in the propagation of uncertainties of the effective sections the neutronic calculation of reactor. Discusses a mini nucleo PWR type according to specifications of the LWR UAM OECD benchmark. Applies the calculation sequence SCALE6.1/TSUNAMI-2D, based on the theory of generalized disturbance (GPT), on the mini nucleo complete, and compared with the result obtained using the two-step method, which combines the GPT with statistical random sampling of the effective sections at the level of the fuel element. Shown that uncertainty in keff increases.

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

    International Nuclear Information System (INIS)

    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 B4C 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. Mechanical

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-01-15

    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.

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

    International Nuclear Information System (INIS)

    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

  12. Consistent Comparison of Full Core PWR Reactivity Initiated Accident with the Method Of Characteristic Code DeCART and the Coarse Mesh Nodal Code PARCS - 180

    International Nuclear Information System (INIS)

    The current state of the art in analysis of a control rod ejection event in a Pressurized Water Reactor (PWR) relies upon the assembly averaged power from a whole core nodal neutronics simulator and some type of pin power reconstruction within the fuel assembly. Recently, there has been interest in taking advantage of the DeCART code to perform a higher fidelity solution which could lead to more accurate pin-power results as well as provide intra-pin power information during the transient. The work described in this paper is the comparison of PARCS and DeCART analysis of two Reactivity Initiated Accidents. The methods used in PARCS and DeCART are briefly described as well as the approach to generate the needed temperature feedbacks. The generation of the macroscopic cross sections and kinetic parameters for PARCS is detailed. The results of both scenarios are shown and the main differences of both approaches are discussed. (authors)

  13. Pulsed Magnetic Welding for Advanced Core and Cladding Steel

    Energy Technology Data Exchange (ETDEWEB)

    Cao, Guoping [Univ. of Wisconsin, Madison, WI (United States); Yang, Yong [Univ. of Florida, Gainesville, FL (United States)

    2013-12-19

    To investigate a solid-state joining method, pulsed magnetic welding (PMW), for welding the advanced core and cladding steels to be used in Generation IV systems, with a specific application for fuel pin end-plug welding. As another alternative solid state welding technique, pulsed magnetic welding (PMW) has not been extensively explored on the advanced steels. The resultant weld can be free from microstructure defects (pores, non-metallic inclusions, segregation of alloying elements). More specifically, the following objectives are to be achieved: 1. To design a suitable welding apparatus fixture, and optimize welding parameters for repeatable and acceptable joining of the fuel pin end-plug. The welding will be evaluated using tensile tests for lap joint weldments and helium leak tests for the fuel pin end-plug; 2 Investigate the microstructural and mechanical properties changes in PMW weldments of proposed advanced core and cladding alloys; 3. Simulate the irradiation effects on the PWM weldments using ion irradiation.

  14. Station blackout core damage frequency in an advanced nuclear reactor

    International Nuclear Information System (INIS)

    Even though nuclear reactors are provided with protection systems so that they can be automatically shut down in the event of a station blackout, the consequences of this event can be severe. This is because many safety systems that are needed for removing residual heat from the core and for maintaining containment integrity, in the majority of the nuclear power plants, are AC dependent. In order to minimize core damage frequency, advanced reactor concepts are being developed with safety systems that use natural forces. This work shows an improvement in the safety of a small nuclear power reactor provided by a passive core residual heat removal system. Station blackout core melt frequencies, with and without this system, are both calculated. The results are also compared with available data in the literature. (author)

  15. Two-stage scaling methodology and direct contact condensation of the core makeup tank in a passive PWR

    International Nuclear Information System (INIS)

    Center for Advanced Reactor Research has developed a passive PWR concept called CARR passive reactor 1300MWe(CP-1300). The CP-1300 has new passive safety features. It is necessary to study the performance of the new safety feature and their interactions in order to assess the response of the CP-1300 under postulated accident conditions. Since it is not feasible to build and test a full power prototype system, a scaled-down integral facility is the best alternative. Therefore, it is necessary to have a rational scaling method. To the end, a two-stage scaling method with validation of scaling methodology is developed for a full-pressure full height scaled-down integral test facility simulating a small-break loss-of-coolant accident in the CP-1300. The present scaling method consists of two stages: scaling methodology stage and validation stage of the scaling methodology and a safety analysis code. In the scaling methodology stage, the whole system and the transient scenario are divided into several subsystems and several phases, respectively. For each subsystem potentially important phenomena are identified and ranked in terms of their relative importance for each phase. The global governing equations for each subsystem are nondimensionalized and expressed as nonlinear integral response functions with nondimensional scaling parameters. In the validation stage of code and scaling methodology the system code is assessed to identify little-known phenomena with high relative importance. The direct contact condensation in the CMT is identified as a little-known phenomenon with high importance. The scaling method is validated through comparison of the key parameters of the model and the prototype using the improved code. For full-height full-pressure scaling it is found out that the power scaling ratio is the same as the volume scaling ratio and the area scaling ratio and the horizontal pipes including the surge line are scaled by an exponent 2/5 of the scaling ratio. If

  16. IRIS - an advanced, grid-appropriate PWR for near-term deployment

    International Nuclear Information System (INIS)

    With the resurgence of nuclear power there is an increasing need for a range of new reactor designs, including smaller units of several hundred MWe. Such reactors fit not only the developing and smaller countries or electric grids, but also provide commercial flexibility to mature markets with large grids by matching the growth, reducing risk, and minimizing financing resources. The International Reactor Innovative and Secure (IRIS) offers an advanced, modular 335 MWe design. IRIS features an integral primary system configuration with all main components located within the reactor vessel. This configuration enables a simplified design with enhanced reliability and economics and supports its safety-by-designTM approach, which results in exceptional safety characteristics. In addition to electricity-only production, IRIS is well suited for cogeneration, including water desalination, district heating, and process steam generation. IRIS is being developed by an international team, led by Westinghouse, incorporating 19 organizations from 10 countries, about half of them European. IRIS development started in 1999 and has reached the level of maturity indicating potential for being commercially offered by the mid of next decade. The preliminary design has been completed and the testing needed for design certification has started last year. The centrepiece of the experimental program is the integral system performance testing to be performed at the SIET facility in Italy. The pre-application review process with the US NRC was initiated in 2002 to address long-lead items, and enable obtaining the Final Design Approval (FDA) by 2013. Economic analyses indicate that IRIS will be competitive with other nuclear and non-nuclear energy sources, whether deployed gradually in single units in smaller grids, or in multiple twin units for larger grids. Additionally, IRIS fits well the recently announced US DOE initiative, GNEP (Global Nuclear Energy Partnership) aiming to support

  17. The integrated PWR

    International Nuclear Information System (INIS)

    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.)

  18. Advanced Technology Application Station Blackout Core Damage Frequency Reduction - The Contribution of an AC Independent Core Residual Heat Removal System

    International Nuclear Information System (INIS)

    An event of station blackout (SBO) can result in severe core damage and undesirable consequences to the public and the environment. To cope with an SBO, nuclear reactors are provided with protection systems that automatically shut down the reactor, and with safety systems to remove the core residual heat. In order to reduce core damage frequency, the design of new reactors incorporates passive systems that rely only on natural forces to operate. This paper presents an evaluation of the SBO core damage frequency of a PWR reactor being designed in Brazil. The reactor has two core residual heat removal systems - an AC dependent system, and a passive system. Probabilistic safety assessment is applied to identify failure scenarios leading to SBO core damage. The SBO is treated as an initiating event, and fault trees are developed to model those systems required to operate in SBO conditions. Event trees are developed to assist in the evaluation of the possible combinations of success or failure of the systems required to cope with an SBO. The evaluation is performed using SAPHIRE, as the software for reliability and risk assessment. It is shown that a substantial reduction in the core damage frequency can be achieved by implementing the passive system proposed for the LABGENE reactor design. Keywords: Station blackout, passive safety system, core damage frequency. (author)

  19. 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)

  20. In-Vessel Retention Technology Development and Use for Advanced PWR Designs in the USA and Korea

    International Nuclear Information System (INIS)

    In-Vessel Retention (IVR) of molten core debris by means of external reactor vessel flooding is a cornerstone of severe accident management for Westinghouse's AP600 (advanced passive light water reactor) design. The case for its effectiveness (made in previous work by the PI) has been thoroughly documented, reviewed as part of the licensing certification, and accepted by the US Nuclear Regulatory Commission. A successful IVR would terminate a severe accident, passively, with the core in a stable, coolable configuration (within the lower head), thus avoiding the largely uncertain accident evolution with the molten debris on the containment floor. This passive plant design has been upgraded by Westinghouse to the AP1000, a 1000 MWe plant very similar to the AP600. The severe accident management approach is very similar too, including In-Vessel Retention as the cornerstone feature, and initial evaluations indicated that this would be feasible at the higher power as well. A similar strategy is adopted in Korea for the APR1400 plant. The overall goal of this project is to provide experimental data and develop the necessary basic understanding so as to allow the robust extension of the AP600 In-Vessel Retention strategy for severe accident management to higher power reactors, and in particular, to the AP1000 advanced passive design

  1. In-Vessel Retention Technology Development and Use for Advanced PWR Designs in the USA and Korea

    Energy Technology Data Exchange (ETDEWEB)

    T.G. Theofanous; S.J. Oh; J.H. Scobel

    2004-05-18

    In-Vessel Retention (IVR) of molten core debris by means of external reactor vessel flooding is a cornerstone of severe accident management for Westinghouse's AP600 (advanced passive light water reactor) design. The case for its effectiveness (made in previous work by the PI) has been thoroughly documented, reviewed as part of the licensing certification, and accepted by the US Nuclear Regulatory Commission. A successful IVR would terminate a severe accident, passively, with the core in a stable, coolable configuration (within the lower head), thus avoiding the largely uncertain accident evolution with the molten debris on the containment floor. This passive plant design has been upgraded by Westinghouse to the AP1000, a 1000 MWe plant very similar to the AP600. The severe accident management approach is very similar too, including In-Vessel Retention as the cornerstone feature, and initial evaluations indicated that this would be feasible at the higher power as well. A similar strategy is adopted in Korea for the APR1400 plant. The overall goal of this project is to provide experimental data and develop the necessary basic understanding so as to allow the robust extension of the AP600 In-Vessel Retention strategy for severe accident management to higher power reactors, and in particular, to the AP1000 advanced passive design.

  2. AP1000® PWR reactor physics analysis with VERA-CS and KENO-VI. Part 1. Zero power physics tests

    International Nuclear Information System (INIS)

    Westinghouse has applied the Core Simulator of the Virtual Environment for Reactor Applications, VERA-CS, under development by the Consortium for Advanced Simulation of LWRs (CASL) to the core physics analysis of the AP1000® PWR. The AP1000 PWR features and advanced first core with radial and axial heterogeneities, including enrichment zoning, multiple burnable absorbers, and a combination of light and heavy control banks to enable the MSHIMTM advanced operational strategy. These advanced features make application of VERA-CS to the AP1000 PWR first core especially relevant to qualify VERA performance. A companion paper at this conference describes the power distribution analysis of the AP1000 PWR with VERA-CS and the KENO Monte-Carlo code. This paper describes the results obtained for the startup physics tests simulations of the AP1000 PWR first core (critical boron, rod worth and reactivity coefficients), supporting the excellent numerical agreement reported in the companion paper for the power distribution. (author)

  3. Advanced neutron source final preconceptual reference core design

    International Nuclear Information System (INIS)

    The preconceptual design phase of the Advanced Neutron Source (ANS) Project ended with the selection of a reference reactor core that will be used to begin conceptual design work. The new reference core consists of two involute fuel elements, of different diameters, aligned axially with a small axial gap between them. The use of different element diameters permits a separate flow of coolant to be provided for each one, thus enhancing the heat removal capability and increasing the thermal-hydraulic margins. The improved cooling allows the elements to be relatively long and thin, so self-shielding is reduced and an acceptable core life can be achieved with a relatively small loading of highly enriched uranium silicide fuel clad in aluminium. The new reference design has a fueled volume 67.4 L, each element having a heated length of 474 mm and a radial fuel thickness of 66 mm. The end-of-cycle peak thermal flux in the large heavy-water reflector tank around the core is estimated to be in the range of 0.8 to 1.0 x 1020 m-2 · s-1. 7 refs., 23 figs., 15 tabs

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

    International Nuclear Information System (INIS)

    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

  5. AP1000® PWR reactor physics analysis with VERA-CS and KENO-VI. Part 2. Power distribution

    International Nuclear Information System (INIS)

    Westinghouse has applied the Core Simulator of the Virtual Environment for Reactor Applications, VERA-CS, under development by the Consortium for Advanced Simulation of LWRs (CASL) to the core physics analysis of the AP1000® PWR. The AP1000 PWR features an advanced first core with radial and axial heterogeneities, including enrichment zoning, multiple burnable absorbers, and a combination of light and heavy control banks to enable the MSHIMTM advanced operational strategy. These advanced features make application of VERA-CS to the AP1000 PWR first core especially relevant to qualify VERA performance. A companion paper at this conference describes the results obtained with VERA-CS and the KENO Monte-Carlo code for startup physics tests simulations of the AP1000 PWR first core (critical boron, rod worth and reactivity coefficients). This paper describes the results of detailed power distribution comparisons between VERA-CS and KENO, and confirms the excellent numerical agreement reported in the companion paper for the startup physics tests simulations. (author)

  6. Advances in in-core and ex-core nuclear instrumentation in the French PWRs

    International Nuclear Information System (INIS)

    Significant advances on both ex-core and in-core instrumentation and protection system have been achieved in France within the last eight years. The most significant achievement is the design, qualification and implementation of the multisection ex-core detectors which are part of the new Integrated Digital Protection System (SPIN) installed on all the 1300 MWe plants in France. Before this, theoretical works and on-reactor tests have been done, starting in 1978 on the BUGEY-2 reactor with a prototype detector, continued in 1981-82 with a protection system mock-up on TRICASTIN-3. These results made it possible to design the new protection system of the 1300 MWe plant program, the first units on operation being PALUEL 1 and 2, which came on line in 1984. During the last years, other tests were made, using fixed in-core instrumentation based on core outlet thermocouples, self-powered detectors and gamma thermometers. The use of this kind of instrumentation is presently being contemplated, probably as a complement of the present instrumentation

  7. Impact of Present Fuel Management Strategies on Maintaining Safety Margins. Mixed Core Aspects. ENUSA's Experience in the PWR Area

    International Nuclear Information System (INIS)

    Due to factors like technology, economic competition, security of supply, etc., the qualification of an alternative nuclear fuel supplier is becoming a more common situation for the nuclear industry. This has contributed to the spreading and generalisation of the Mixed Cores configurations, i.e. those cores containing fuel assemblies of different designs, either from the same or different Fuel Manufacturer. The generalisation of the mixed-cores situation has led to the implementation at ENUSA of a systematic design process aimed to the evaluation and licensing of the mixed-cores configurations and the individual aspects of each different fuel design which is involved. Furthermore, with the in-house design capabilities and the introduction of improved design methodologies, the fuel assembly designs are addressed by ENUSA from their inception with the basic design objective of minimising the potential unfavourable mixed core effects affecting the safety margins. Naturally the scope and level of the required analyses depends on the level of fuel design changes and methodologies that are introduced. The basic objective that is pursued is the integral assessment of all the design areas in order to demonstrate the compliance with the licensing basis of each particular application, not impacting or even improving the current plant safety margins or improving them when possible. This paper reviews the compatibility assessment aspects, the development and design verification process, the structural performance, and the hydraulic and thermal-hydraulic performance

  8. 3D neutron transport and HPC. A PWR full core calculation using PENTRAN SN code and IBM BLUEGENE/P computers

    International Nuclear Information System (INIS)

    When dealing with nuclear reactor calculation schemes, the need for 3D transport-based reference solutions is essential for validation and optimization purposes. As SN transport method may be considered promising with respect to comprehensive parallel computations, a 3D full PWR core benchmark was proposed to challenge the capabilities of the PENTRAN parallel SN code utilizing an IBM-BG/P computer. After a brief description of the benchmark, a parallel performance analysis is carried out, and shows that the parallelizable (Amdahl) fraction of PENTRAN is comprised between 0.994 ≤ f ≤ 0.996 for a number of BG/P nodes ranging from 17 to 1156. The associated speedup reaches a value greater than 200 with 1156 nodes. Using a best estimate model, PENTRAN results are then compared to Monte Carlo results rendered using the MCNP5 code. Good consistency is observed between the two methods (SN and Monte Carlo), with discrepancies less than 65 pcm for the keff, and less than 2.5% for the flux at the pincell level. (author)

  9. 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.

  10. Development and assessment of advanced reactor core protection system

    International Nuclear Information System (INIS)

    An advanced core protection system for a pressurized water reactor, Reactor Core Protection System (RCOPS), was developed by adopting a high performance hardware platform and optimal system configuration. The functional algorithms of the core protection system were also improved to enhance the plant availability by reducing unnecessary reactor trips and increasing operational margin. The RCOPS consists of four independent safety channels providing a two-out-of-four trip logic. The reliability analysis using the reliability block diagram method showed the unavailability of the RCOPS to be lower than the conventional system. The failure mode and effects analysis demonstrated that the RCOPS does not lose its intended safety functions for most failures. New algorithms for the RCOPS functional design were implemented in order to avoid unnecessary reactor trips by providing auxiliary pre-trip alarms and signal validation logic for the control rod position. The new algorithms in the RCOPS were verified by comparing the RCOPS calculations with reference results. The new thermal margin algorithm for the RCOPS was expected to increase the operational margin to the limit for Departure from Nucleate Boiling Ratio (DNBR) by approximately 1%. (author)

  11. Design Requirements of an Advanced HANARO Reactor Core Cooling System

    International Nuclear Information System (INIS)

    An advanced HANARO Reactor (AHR) is an open-tank-type and generates thermal power of 20 MW and is under conceptual design phase for developing it. The thermal power is including a core fission heat, a temporary stored fuel heat in the pool, a pump heat and a neutron reflecting heat in the reflector vessel of the reactor. In order to remove the heat load, the reactor core cooling system is composed of a primary cooling system, a primary cooling water purification system and a reflector cooling system. The primary cooling system must remove the heat load including the core fission heat, the temporary stored fuel heat in the pool and the pump heat. The purification system must maintain the quality of the primary cooling water. And the reflector cooling system must remove the neutron reflecting heat in the reflector vessel of the reactor and maintain the quality of the reflector. In this study, the design requirement of each system has been carried out using a design methodology of the HANARO within a permissible range of safety. And those requirements are written by english intend to use design data for exporting the research reactor

  12. Closed-analytical mixing model describing the slug transport from the inlet nozzle to the reactor core in PWR

    International Nuclear Information System (INIS)

    A closed-analytical model describing the slug transport from the inlet nozzle to the reactor core is presented. For the time being, it is applicable only to steady flow and probably only for symmetrical flow conditions. Using an analytical solution of the 1D diffusion equation for linear time-variable diffusion coefficients found by the authors, mixing in a transient flow may be modelled as well. (orig.)

  13. Mass-flow measurements under PWR reflood conditions in a downcomer and at a core barrel vent-valve location

    International Nuclear Information System (INIS)

    Instrumentation schemes have been devised and calibrated for two-phase mass flux measurement under simulated pressurized-water reactor reflood conditions. The instrumentation schemes, consisting of two instruments, are to be located in a downcomer and at a vent valve location. More specifically, a drag disk and string probe package will be positioned in three locations in the Japanese Slab Core Test Facility downcomer; a turbine meter and string probe combination will be placed in the core barrel vent valve region of the Japanese Cylindrical Core Test Facility. Calibration in air/water flow under steady-state conditions yielded mass flux algorithms for both instrumentation schemes. The drag disk and string probe calibration gave mass flux estimates to within +40%, -30%. The turbine meter and string probe combination produced estimates to within +-30% for high mass fluxes and +-70% at the low flow rates. Considering the chaotic nature of the flow, the wide range of mass fluxes measured, and the simple homogeneous models employed, the algorithms predict mass flux values reasonably well

  14. Application of diffusion theory methods to PWR [pressurized water reactors] analysis

    International Nuclear Information System (INIS)

    In-core physics analysis of pressurized light water reactors (PWRs) requires accurate predictions of three-dimensional pin-by-pin power distributions. The PWR analyses must rely on diffusion theory approximation because no practical methods exist for performing routine three-dimensional pin-by-pin transport calculations. Pin-by-pin diffusion calculations are also prohibitively expensive in three-dimensional geometry, and PWR analyses utilize either two-dimensional pin-by-pin models or three-dimensional advanced nodal models. The purpose of this paper is to detail and contrast approximations required by pin-by-pin and nodal diffusion methods

  15. Severe accidents: the primary and secondary bleed and feed procedures to prevent PWR high pressure core melting

    International Nuclear Information System (INIS)

    New strategies to cope with severe reactor accidents leading to core degradation and eventually to a release of large quantities of radioactive products into the environment, have been developed in several countries over the last few years. In order to reduce the probability and risk associated with such grave events, appropriate accident management measures have been defined. The most interesting procedures for the prevention of an irreversible deterioration of the plant status and the maintenance of the core in coolable conditions are the secondary and primary side Bleed and Feed measures. In fact, in case of loss of secondary heat sink accidents, these procedures aim either to recover the secondary side heat removal capability by depressurization of the steam generators or to remove the residual heat via the pressurizer valves. In this way the probability of core meltdown with high primary pressure is drastically reduced. Recent investigations on primary and secondary side Bleed and Feed interventions have already shown the high potentiality of this kind of actions in using systems such as feedwater storage tank, accumulators, fire fighting systems or mobile pumps. Since the use of these procedures is strongly dependent on the intervention logic and on the characteristics of the specific plant design, there is the need of carrying out accurate analyses to assess and optimize the intervention actions. This report sets itself a goal in defining a basis for the study of transients which can be handled with Bleed and Feed procedures, allowing in this respect future analyses of the Swiss PWRs. (author) 6 figs., 15 refs

  16. Estimation of core-damage frequency to evolutionary ALWR [advanced light water reactor] due to seismic initiating events: Task 4.3.3

    International Nuclear Information System (INIS)

    The Electric Power Research Institute (EPRI) is presently developing a requirements document for the design of advanced light water reactors (ALWRs). One of the basic goals of the EPRI ALWR Requirements Document is that the core-damage frequency for an ALWR shall be less than 1.0E-5. To aid in this effort, the Department of Energy's Advanced Reactor Severe Accident Program (ARSAP) initiated a functional probabilistic risk assessment (PRA) to determine how effectively the evolutionary plant requirements contained in the existing EPRI Requirements Document assure that this safety goal will be met. This report develops an approximation of the core-damage frequency due to seismic events for both evolutionary plant designs (pressurized-water reactor (PWR) and boiling-water reactor(BWR)) as modeled in the corresponding functional PRAs. Component fragility values were taken directly form information which has been submitted for inclusion in Appendix A to Volume 1 of the EPRI Requirements Document. The results show a seismic core-damage frequency of 5.2E-6 for PWRS and 5.0E-6 for BWRs. Combined with the internal initiators from the functional PRAs, the overall core-damage frequencies are 6.0E-6 for the pwr and BWR, both of which satisfy the 1.0E-5 EPRI goal. In addition, site-specific considerations, such as more rigid components and less conservative fragility data and seismic hazard curves, may further reduce these frequencies. The effect of seismic events on structures are not addressed in this generic evaluation and should be addressed separately on a design-specific basis. 7 refs., 6 figs., 3 tabs

  17. A comparative analysis of behaviour of PWR reactor cores in dangerous transient regimes by using the catastrophe theory

    International Nuclear Information System (INIS)

    The paper reports on the transitory regimes that occur within the core of CANDU and WWER reactor regimes which might initiate boiling possibly resulting in local structural element failures. Neutron, thermal and hydraulic phenomena intermingle each other by feedback and can get out of control. The time variation of vapour bubble size and density as a major random phenomenon affects essentially the reactor stability. The study of this phenomenon have been performed on the basis of a statistical mathematical model. The analysis has revealed that the most critical situations occur directly after the initiation of the transient regimes and consequently the critical ranges of variations should be avoided. A greater temperature gradient inside the boundary layer is required in order to ensure the vapour bubble breaking. The mathematical model developed in this paper allows the study of the system behaviour for a large range of thermohydraulic and geometrical quantities. (authors). 16 figs., 7 refs

  18. 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)

    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

  19. Comparison of advanced mid-sized reactors regarding passive features, core damage frequencies and core melt retention features

    International Nuclear Information System (INIS)

    New Light Water Reactors, whose regular safety systems are complemented by passive safety systems, are ready for the market. The special aspect of passive safety features is their actuation and functioning independent of the operator. They add significantly to reduce the core damage frequency (CDF) since the operator continues to play its independent role in actuating the regular safety devices based on modern instrumentation and control (I and C). The latter also has passive features regarding the prevention of accidents. Two reactors with significant passive features that are presently offered on the market are the AP1000 PWR and the SWR 1000 BWR. Their passive features are compared and also their core damage frequencies (CDF). The latter are also compared with those of a VVER-1000. A further discussion about the two passive plants concerns their mitigating features for severe accidents. Regarding core-melt retention both rely on in-vessel cooling of the melt. The new VVER-1000 reactor, on the other hand features a validated ex-vessel concept. (author)

  20. Modeling and interpretation of the physical phenomena inducing noise on PWR's ex-core and in-core detectors signals. Potentialities of an on-line surveillance based on this noise measurement and analysis

    International Nuclear Information System (INIS)

    The study of the 900 MW PWR's ex-core neutron sensors signal fluctuations and their interpretation were to be carried out in this work. The aim, at the end was to be able to decide what could be the possibilities and the limits of a surveillance based on the pressurized water reactors neuton noise analysis. Hence, all the perturbation sources whether they are currently present or just may occur, were to be studied under their two aspects: qualitative and quantitative. The small amplitudes of the perturbations to be considered allowed their effect on the neutron detector signals to be calculated with the first order approximation of the perturbation theory model. We developed the model and the computer software which carried out the calculation of the detectors signal sensitivity to the perturbations that were considered. Moreover, our work was supported by numerous on-site measurements, carried out on many reactors. The following facts can be extracted from the work: - neutron noise analysis allows the vibrational behavior characterization of some primary circuit main components, at locations where no direct vibratory measurements can be generally considered, - some signatures corresponding to unsuitable behaviors were obtained, - other signatures typical of unsuitable behaviors, though not observed, have been defined, - besides the vibratory area, neutron noise may be used for the characterization of the detector state itself, - in the thermohydraulical field and for PWRs, the neutron noise possibilities have to be considered as rather low though not zero

  1. 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)

    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 233U, 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 233U, 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 ThPuO2 fuel, and the second is with standard or high MR and ThUO2 fuel. The first way led to the design of under-moderated HCPWR following the criteria of high 233U 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 233U 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 233U production is the limiting factor. That is why it was eventually proposed to study how the production of 233U within PWR (with standard MR), from 2020. It was shown that the anticipated production of 233U 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)

  2. Advanced light microscopy core facilities: Balancing service, science and career.

    Science.gov (United States)

    Ferrando-May, Elisa; Hartmann, Hella; Reymann, Jürgen; Ansari, Nariman; Utz, Nadine; Fried, Hans-Ulrich; Kukat, Christian; Peychl, Jan; Liebig, Christian; Terjung, Stefan; Laketa, Vibor; Sporbert, Anje; Weidtkamp-Peters, Stefanie; Schauss, Astrid; Zuschratter, Werner; Avilov, Sergiy

    2016-06-01

    Core Facilities (CF) for advanced light microscopy (ALM) have become indispensable support units for research in the life sciences. Their organizational structure and technical characteristics are quite diverse, although the tasks they pursue and the services they offer are similar. Therefore, throughout Europe, scientists from ALM-CFs are forming networks to promote interactions and discuss best practice models. Here, we present recommendations for ALM-CF operations elaborated by the workgroups of the German network of ALM-CFs, German Bio-Imaging (GerBI). We address technical aspects of CF planning and instrument maintainance, give advice on the organization and management of an ALM-CF, propose a scheme for the training of CF users, and provide an overview of current resources for image processing and analysis. Further, we elaborate on the new challenges and opportunities for professional development and careers created by CFs. While some information specifically refers to the German academic system, most of the content of this article is of general interest for CFs in the life sciences. Microsc. Res. Tech. 79:463-479, 2016. © 2016 THE AUTHORS MICROSCOPY RESEARCH AND TECHNIQUE PUBLISHED BY WILEY PERIODICALS, INC. PMID:27040755

  3. 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)

    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

  4. Verification of the Advanced Nodal Method on BWR Core Analyses by Whole-Core Heterogeneous Transport Calculations

    International Nuclear Information System (INIS)

    Recent boiling water reactor (BWR) core and fuel designs have become more sophisticated and heterogeneous to improve fuel cycle cost, thermal margin, etc. These improvements, however, tend to lead to a strong interference effect among fuel assemblies, and it my cause some inaccuracies in the BWR core analyses by advanced nodal codes. Furthermore, the introduction of mixed-oxide (MOX) fuel will lead to a much stronger interference effect between MOX and UO2 fuel assemblies. However, the CHAPLET multiassembly characteristics transport code was developed recently to solve two-dimensional cell-heterogeneous whole-core problems efficiently, and its results can be used as reference whole-core solutions to verify the accuracy of nodal core calculations. In this paper, the results of nodal core calculations were compared with their reference whole-core transport solutions to verify their accuracy (in keff, assembly power and pin power via pin power reconstruction) of the advanced nodal method on both UO2 and MOX BWR whole-core analyses. Especially, it was investigated if there were any significant differences in the accuracy between MOX and UO2 results

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

    International Nuclear Information System (INIS)

    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 UO2–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

  6. Reflooding phase after loss of coolant of an advanced pressurized water reactor with high conversion ratio

    International Nuclear Information System (INIS)

    The emergency core cooling behaviour of an advanced pressurized water reactor (APWR) during the reflooding phase of the LOCA with double-ended break is analysed and compared to a common pressurized water reactor (PWR). The code FLUT-BS, its models and correlations are explained in detail and have been verified by numerous PWR-reflood experiments with large parameter range. The influence of core-design on ECC-behaviour as well as the influences of initial and boundary values are examined. The results show the essential differences of ECC-behaviour between PWR and APWR. (orig.)

  7. An advancement in iterative solution schemes for three-dimensional, two-fluid modeling of two-phase flow in PWR fuel bundles

    International Nuclear Information System (INIS)

    Highlights: • A fully three-dimensional two-fluid model coupled with heat conduction was outlined. • Two-fluid numerical scheme capability was evaluated against NUPEC PSBT Benchmark. • GMRES, FGMRES, DQGMRES, CGNR, BCG, and TFQMR solvers were tested as iterative schemes. • Candidate Krylov solvers do not introduce deviations to the two-phase flow results. • GMRES, FGMRES, and DQGMRES have a more efficient and stable convergence performance. - Abstract: This paper outlines a fully three-dimensional two-fluid one-pressure model with a semi-implicit finite difference scheme coupled with heat conduction which can be applicable to thermal non-equilibrium two-phase flow field in subchannel geometry of Pressurized Water Reactors (PWR). The system of equations was linearized using the Newton–Raphson method and was collapsed into the pressure equations forming a system of the Poisson type. Then, two-phase flow modeling was combined with Krylov methods as advanced computing techniques to investigate the feasibility of implementing preconditioned Krylov subspace solvers as the numerical scheme to solve pressure equations. Six popular Krylov subspace solvers were considered: GMRES, FGMRES, DQGMRES, CGNR, BCG, and TFQMR combined with the block incomplete LU factorization with a dual truncation strategy (BILUT) preconditioner. These proposed iterative solvers were applied to the constructed linear pressure equations in the inner iteration in combination with the outer-Raphson iteration loop. Evaluation was performed in two stages. First, two-fluid numerical scheme capability was evaluated against OECD/NRC NUPEC PWR Bundle tests (PSBT Benchmark). The results for steady-state (PSBT) bundle show that an overall agreement can be found. At the second stage, convergency, stability, and accuracy of the proposed schemes were studied based on PSBT steady-state data through a comparison of utilized Krylov solvers and the direct inversion method as the pressure solution

  8. The integrated code system CASCADE-3D for advanced core design and safety analysis

    International Nuclear Information System (INIS)

    The new program system CASCADE-3D (Core Analysis and Safety Codes for Advanced Design Evaluation) links some of Siemens advanced code packages for in-core fuel management and accident analysis: SAV95, PANBOX/COBRA and RELAP5. Consequently by using CASCADE-3D the potential of modern fuel assemblies and in-core fuel management strategies can be much better utilized because safety margins which had been reduced due to conservative methods are now predicted more accurately. By this innovative code system the customers can now take full advantage of the recent progress in fuel assembly design and in-core fuel management.(author)

  9. Transition core DNBR penalty determination for Angra-1 nuclear power plant mixed cores consisting of standard and advanced fuel assemblies

    International Nuclear Information System (INIS)

    When two (or more) types of Fuel Assemblies (FA) are inserted in a nuclear reactor core, a flow redistribution occurs, due to the different hydraulic resistances of these assemblies. This way, the FA's with higher hydraulic resistance will get a Departure from Nucleate Boiling Ratio (DNBR) penalty since a part of the total flow will diverge to the FA's with lower hydraulic resistance. Regarding Angra-1 Nuclear Power Plant (NPP), it is planned in a next cycle to insert a new Advanced FA that is a result from a joint-venture project of the companies INB - Industrias Nucleares do Brasil, WEC - Westinghouse Electric Company and KNF - Korean Nuclear Fuel. Therefore, the purpose of this article is to show the work done to determine the DNBR penalty to be applied to the Advanced FA's present in a mixed (or transition) core consisting of Advanced and Standard FA's. (author)

  10. Conceptual studies of core catchers for advanced LWRs

    International Nuclear Information System (INIS)

    In the present work, we first identify thoroughly the requirements, which an ex-vessel molten core retention system/strategy has to fulfil in order to cope with the most violent phenomena, following a postulated Reactor Pressure Vessel (RPV) failure. The state of the art is briefly reviewed with particular emphasis on the question to what extent the currently proposed concepts could meet the stated requirements, taken as a whole. Two core-catcher concepts, resulting from two distinct applications of a 'realistic' approach are described altogether with generic R and D works needed for a further development of such devices. Examples of preliminary design calculations and experimental facilities under construction are given for the purpose of illustrating the progress status of the ex-vessel molten core retention research program at CEA/DRN. (author)

  11. Applying CLSM to increment core surfaces for histometric analyses: A novel advance in quantitative wood anatomy

    OpenAIRE

    Wei Liang; Ingo Heinrich; Gerhard Helle; I. Dorado Liñán; T. Heinken

    2013-01-01

    A novel procedure has been developed to conduct cell structure measurements on increment core samples of conifers. The procedure combines readily available hardware and software equipment. The essential part of the procedure is the application of a confocal laser scanning microscope (CLSM) which captures images directly from increment cores surfaced with the advanced WSL core-microtome. Cell wall and lumen are displayed with a strong contrast due to the monochrome black and green nature of th...

  12. Development of hybrid core calculation system using two-dimensional full-core heterogeneous transport calculation and three-dimensional advanced nodal calculation

    International Nuclear Information System (INIS)

    This paper presents a description of the Hybrid Core Calculation System which is based on a very rigorous but practical method utilizing best estimate core design calculations and taking advantage of the recent remarkable progress of computers. The basic idea of this system is to generate the correction factors for assembly-homogenized cross sections, discontinuity factors, etc. by comparing the CASMO-4 and SIMULATE-3 2-D full-core calculation results under the consistent calculation condition and applying them to the SIMULATE-3 3-D calculation. The CASMO-4 2-D heterogeneous core calculation is performed for each depletion step using the core conditions previously determined by ordinary SIMULATE-3 core calculations. This avoids time-consuming iterative calculations of the critical boron concentration search and the thermal hydraulic feedback. These calculations are instead performed in the final SIMULATE-3 3-D calculation using the previously determined correction factors. The Hybrid Core Calculation System was verified using data from a commercial PWR for several cycles, and it was demonstrated that the accuracy of core calculation is improved. (author)

  13. Phebus-SFD B9+ experiment on the degradation of a PWR Type Core. Comparison report. Volume 1 + 2. ISP 28 OECD/NEA/CSNI International standard problem n. 28

    International Nuclear Information System (INIS)

    The PHEBUS Severe Fuel Damage B9+ test was accepted by OECD/CSNI as the International Standard Problem No. 28. The aim of the exercise was to access the ability of the codes to describe the degradation of PWR-type cores and in particular to predict the main phenomena observed in B9+: cladding oxidation, H2 production, fuel dissolution by molten Zr and relocation of melts resulting of chemical interaction. Fifteen calculations performed with seven different codes were submitted in semi-blind conditions. Code to data comparisons enabled the current ability of the codes to be identified and to point out weaknesses and lacks regarding the UO2 dissolution and melt relocation. Finally, this exercise emphasized the need to improve and complete the modelling of the main degradation processes and to better assess these models to be used for reactor calculations

  14. Core loading pattern optimization of a typical two-loop 300 MWe PWR using Simulated Annealing (SA), novel crossover Genetic Algorithms (GA) and hybrid GA(SA) schemes

    International Nuclear Information System (INIS)

    Highlights: • SA and GA based optimization for loading pattern has been carried out. • The LEOPARD and MCRAC codes for a typical PWR have been used. • At high annealing rates, the SA shows premature convergence. • Then novel crossover and mutation operators are proposed in this work. • Genetic Algorithms exhibit stagnation for small population sizes. - Abstract: A comparative study of the Simulated Annealing and Genetic Algorithms based optimization of loading pattern with power profile flattening as the goal, has been carried out using the LEOPARD and MCRAC neutronic codes, for a typical 300 MWe PWR. At high annealing rates, Simulated Annealing exhibited tendency towards premature convergence while at low annealing rates, it failed to converge to global minimum. The new ‘batch composition preserving’ Genetic Algorithms with novel crossover and mutation operators are proposed in this work which, consistent with the earlier findings (Yamamoto, 1997), for small population size, require comparable computational effort to Simulated Annealing with medium annealing rates. However, Genetic Algorithms exhibit stagnation for small population size. A hybrid Genetic Algorithms (Simulated Annealing) scheme is proposed that utilizes inner Simulated Annealing layer for further evolution of population at stagnation point. The hybrid scheme has been found to escape stagnation in bcp Genetic Algorithms and converge to the global minima with about 51% more computational effort for small population sizes

  15. A study on the advanced statistical core thermal design methodology

    International Nuclear Information System (INIS)

    A statistical core thermal design methodology for generating the limit DNBR and the nominal DNBR is proposed and used in assessing the best-estimate thermal margin in a reactor core. Firstly, the Latin Hypercube Sampling Method instead of the conventional Experimental Design Technique is utilized as an input sampling method for a regression analysis to evaluate its sampling efficiency. Secondly and as a main topic, the Modified Latin Hypercube Sampling and the Hypothesis Test Statistics method is proposed as a substitute for the current statistical core thermal design method. This new methodology adopts 'a Modified Latin Hypercube Sampling Method' which uses the mean values of each interval of input variables instead of random values to avoid the extreme cases that arise in the tail areas of some parameters. Next, the independence between the input variables is verified through 'Correlation Coefficient Test' for statistical treatment of their uncertainties. And the distribution type of DNBR response is determined though 'Goodness of Fit Test'. Finally, the limit DNBR with one-sided 95% probability and 95% confidence level, DNBR95/95' is estimated. The advantage of this methodology over the conventional statistical method using Response Surface and Monte Carlo simulation technique lies in its simplicity of the analysis procedure, while maintaining the same level of confidence in the limit DNBR result. This methodology is applied to the two cases of DNBR margin calculation. The first case is the application to the determination of the limit DNBR where the DNBR margin is determined by the difference between the nominal DNBR and the limit DNBR. The second case is the application to the determination of the nominal DNBR where the DNBR margin is determined by the difference between the lower limit value of the nominal DNBR and the CHF correlation limit being used. From this study, it is deduced that the proposed methodology gives a good agreement in the DNBR results with

  16. Analysis of core calculation schemes for advanced water reactors

    International Nuclear Information System (INIS)

    This research thesis addresses the analysis of the core control of sub-moderated water reactors with plutonium fuel and varying spectrum. Firstly, a calculation scheme is defined, based on transport theory for the three existing assembly configurations. It is based on the efficiency analysis of the control cluster and of the flow sheet shape in the assembly. Secondly, studies of the assembly with control cluster and within a theory of diffusion with homogenization or detailed assembly representation are performed by taking the environment into account in order to assess errors. Thirdly, due to the presence of a very efficient absorbent in control clusters, a deeper physical analysis requires the study of the flow gradient existing at the interface between assemblies. A parameter is defined to assess this gradient, and theoretically calculated by using finite elements. Developed software is validated

  17. Advanced neutron source three-element-core fuel grading

    International Nuclear Information System (INIS)

    The proposed advanced neutron source (ANS) neutron research facility's purpose is to provide unprecedented experimental capabilities in the areas of neutron scattering, materials research, and isotope production. The primary goals of the ANS project are to obtain neutron flux levels that are 5 to 10 times larger than any current existing facility and to provide isotope irradiation facilities that are at least as good as the High-Flux Isotope Reactor at Oak Ridge National Laboratory. The design changes in the ANS are described

  18. Verification of advanced methods in TARMS boiling water reactor core management system

    International Nuclear Information System (INIS)

    The TARMS (Toshiba Advanced Reactor Management System) software package was developed as an effective on-line, on-site Boiling Water Reactor (BWR) core operation management system. It covers almost all the functional requirements to the current process computer to increase on-site core management capability, capacity factors, thermal margins, fuel reliability, and so on, by supporting application functions for monitoring the present core power distribution, and for aiding site engineers in making the core operation plans, by predicting future core performance. It is based on a three dimensional, 1.5 energy group, coarse mesh nodal diffusion theory code ''LOGOS02'', and includes advanced methods to increase the accuracy of core power distribution calculations as well as a local peaking factor calculation method by which the effect of neighboring nodes on intra-nodal power distribution can be considered. TARMS has been installed in eight BWR plants and was verified to be an effective BWR core operation management tool. This paper describes its advanced methods and the results of verifications with actual plant data. (author). 3 refs, 6 figs

  19. An Advanced Diagnostic Display for Core Protection Calculator System

    International Nuclear Information System (INIS)

    The main purpose of a Nuclear Power Plant Instrumentation and Control (I and C) Display System is to provide operator's interface for I and C systems. The CPCS display(Shin-Kori 1 and 2) provides operators with 1) plant monitoring values of field input and algorithm variables that reflect the reactor core conditions, 2) operation values that operators can change and 3) CPCS status. It will be an optimal case if operators can understand the plant (including CPCS itself) condition intuitively with the displayed values but it is not easy in CPCS. For example, if the CPCS Channel Trouble light is lit, operators need some amount of time to investigate what caused the trouble light because there are more than hundred causes that can generate the channel trouble. If a Display supports diagnostic information that shows what cause the displayed alarms, it will greatly help operators in easy understanding the CPCS status. To provide these diagnostic information, this paper suggests an active self-explanatory display mechanism. This self-explanatory diagnostic display mechanism utilizes an ontology in XML that describes parent child, sibling relationships of display variables, through which in-depth, in-breadth diagnostic tracking is possible. This paper consists of two parts. First, the key features of CPCS Flat Panel Display System (FPDS) are described. Second, the features of active self explanatory diagnostic display are discussed

  20. Minor actinide transmutation on PWR burnable poison rods

    International Nuclear Information System (INIS)

    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 keff 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 keff markedly. The PWR keff 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

  1. Operation flexibility and availability improvements using BEACON, an advanced core monitoring system

    International Nuclear Information System (INIS)

    In response to utilities needs in improving plant operation flexibility and plant availability, Westinghouse introduced the advanced core monitoring and operational support system, BEACON, two years ago. Since then, the continuous development of the BEACON system has led to significant advances in further reducing utilities Operation and Maintenance (O and M) costs. The development of the BEACON system is made possible by two breakthroughs: 1) advanced numerical method to solve the diffusion equations extremely fast and 2) development of cost effective, state-of-the-art computing system, workstation. This paper presents the numerical scheme used in the neutronic solution and how BEACON uses the core instrumentations to provide the continuous three-dimensional (3D) core power distribution. Once the state of the core is known on a continuous basis, several indirect surveillance and/or Technical Specifications on core power distribution can be relaxed or totally eliminated. Section 1 outlines the numerical scheme used in BEACON for solving the diffusion equations and to provide the 3D continuous power distribution. Section 2 describes the hardware requirements. Section 3 discusses applications of BEACON to improve plant operation flexibility and plant availability. Examples of actual BEACON usage to demonstrate its effectiveness are presented in Section 4 and the paper is closed with a summary of future directions. (author). 4 refs, 6 figs

  2. Advanced calculational methods for power reactors and LWR core design parameters

    International Nuclear Information System (INIS)

    The purpose of the Specialists Meeting on Advanced Calculational Methods for Power Reactors, held in Cadarache, France, 10-14 September 1990, was to provide a forum for reviewing and discussing selected core physics of water cooled reactors (including high convertors). New methods of advanced calculation for advanced fuels and complex geometries of next generation reactors with a high level of accuracy were discussed and the importance of supercomputing and on-line monitoring was also acknowledged. The meeting was attended by about 60 participants from 20 countries who presented 30 papers. The Technical Committee Meeting on LWR Core Design Parameters, held in Rez, former Czechoslovakia, 7-11 October 1991, provided an opportunity for participants to exchange their experience on reactor physics aspects of benchmark calculations of various lattices, methods for core parameter calculations, core monitoring and in-core fuel management. At the Workshop there were further discussions related to the benchmark problems, homogenization techniques and cross-section representations. Thirty-five papers were presented by about 43 participants from 19 countries. A separate abstract was prepared for each of the mentioned papers. Refs, figs and tabs

  3. The Importance of Hosting a Codex Committee in Advancing Codex's Strategic Vision and Core Values

    OpenAIRE

    Backhouse, Ann

    2014-01-01

    The role the host country of a Codex Committee plays in promoting participation in the setting of international food standards as well as advancing the Codex agenda with regards to transparency and inclusiveness is an important one. Of particluar significance is the obligation of Codex members to promote the strategic vision of Codex and its core values in the undertaking of this leadership role

  4. BEACON - An advanced continuous core monitoring and operational support system for pressurized water reactors

    International Nuclear Information System (INIS)

    An advanced continuous core monitoring and operational support system, BEACON, has been developed which combines a super fast nodal model, workstation based hardware, and existing instrumentation which can be used to improve plant availability and operating margin. (author). 6 refs, 8 figs

  5. Validation of BWR advanced core and fuel nuclear designs with power reactor measurements

    International Nuclear Information System (INIS)

    Power reactor measurements have been important in validating the reliability, performance characteristics and economics of BWR advanced core and fuel designs. Such measurements go beyond the data obtainable from normal reactor operation and provide detailed benchmark data necessary to verify design and licensing computer design and simulation models. In some cases, such as in the validation of the performance of zirconium barrier pellet-cladding-interaction (PCI) resistant cladding, the BWR power reactor measurements have subjected the advanced fuel design to operating conditions more severe than normal operating conditions, thereby providing nuclear-thermal-mechanical-corrosion performance data for accelerated or extended conditions of operation. In some cases destructive measurements have been carried out on BWR power reactor fuel to provide microscopic and macroscopic data of importance in validating design and licensing analysis methods. There is not uniform agreement among core and fuel designers on the needs for special power reactor core and fuel measurements for validation of advanced designs. The General Electric approach has been to error on the side of extensive, detailed measurements so as to assure reliable performance licensing and economic design and predictive capability. This paper is a summary of some of the validative power reactor measurements that have been carried out on advanced BWR core and fuel designs. Some comparisons of predictions with the data are summarized

  6. Chemical System Decontamination at PWR Power Stations Biblis A and B by Advanced System Decontamination by Oxidizing Chemistry (ASDOC-D) Process Technology - 13081

    International Nuclear Information System (INIS)

    For chemical decontamination of PWR primary systems the so called ASDOC-D process has been developed and qualified at the German PWR power station Biblis. In comparison to other chemical decontamination processes ASDOC-D offers a number of advantages: - ASDOC-D does not require separate process equipment but is completely operated and controlled by the nuclear site installations. Feeding of chemical concentrates into the primary system is done by means of the site's dosing systems. Process control is performed by standard site instrumentation and analytics. - ASDOC-D safely prevents any formation and precipitation of insoluble constituents - Since ASDOC-D is operated without external equipment there is no need for installation of such equipment in high radioactive radiation surrounding. The radioactive exposure rate during process implementation and process performance may therefore be neglected in comparison to other chemical decontamination processes. - ASDOC-D does not require auxiliary hose connections which usually bear high leakage risk. The above mentioned technical advantages of ASDOC-D together with its cost-effectiveness gave rise to Biblis Power station to agree on testing ASDOC-D at the volume control system of PWR Biblis unit A. By involving the licensing authorities as well as expert examiners into this test ASDOC-D received the official qualification for primary system decontamination in German PWR. As a main outcome of the achieved results NIS received contracts for full primary system decontamination of both units Biblis A and B (each 1.200 MW) by end of 2012. (authors)

  7. Qualification according to PDI's techniques UT EPRI methodology Phased Array for the inspection of vessels of PWR reactor with advanced robotic equipment

    International Nuclear Information System (INIS)

    The techniques and procedures qualified in the program EPRI PDI are directly applicable in plants whose reference code is ASME XI - specifically the Appendix VIII-, mainly USA and countries in which it is established American PWR technology. While countries with reactors in operation technology ABB (Sweden) or type VVER (Finland and Eastern countries) requires a qualification of specific technical type ENIQ, PDI qualification is a valuable reference since it allows to deal with such qualifications with guarantees. (Author)

  8. Core and Refueling Design Studies for the Advanced High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Ilas, Dan [ORNL; Varma, Venugopal Koikal [ORNL; Cisneros, Anselmo T [ORNL; Kelly, Ryan P [ORNL; Gehin, Jess C [ORNL

    2011-09-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a central generating station type [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. This report presents the current status of ongoing design studies of the core, in-vessel structures, and refueling options for the AHTR. The AHTR design remains at the notional level of maturity as important material, structural, neutronic, and hydraulic issues remain to be addressed. The present design space exploration, however, indicates that reasonable options exist for the AHTR core, primary heat transport path, and fuel cycle provided that materials and systems technologies develop as anticipated. An illustration of the current AHTR core, reactor vessel, and nearby structures is shown in Fig. ES1. The AHTR core design concept is based upon 252 hexagonal, plate fuel assemblies configured to form a roughly cylindrical core. The core has a fueled height of 5.5 m with 25 cm of reflector above and below the core. The fuel assembly hexagons are {approx}45 cm across the flats. Each fuel assembly contains 18 plates that are 23.9 cm wide and 2.55 cm thick. The reactor vessel has an exterior diameter of 10.48 m and a height of 17.7 m. A row of replaceable graphite reflector prismatic blocks surrounds the core radially. A more complete reactor configuration description is provided in Section 2 of this report. The AHTR core design space exploration was performed under a set of constraints. Only low enrichment (<20%) uranium fuel was considered. The coated particle fuel and matrix materials were derived from those being developed and demonstrated under the Department of Energy Office of Nuclear Energy (DOE-NE) advanced gas reactor program. The coated particle volumetric packing fraction was restricted to at most 40%. The pressure

  9. Training reactor deployment. Advanced experimental course on designing new reactor cores

    International Nuclear Information System (INIS)

    Czech Technical University in Prague (CTU) operating its training nuclear reactor VR1, in cooperation with the North West University of South Africa (NWU), is applying for accreditation of the experimental training course ''Advanced experimental course on designing the new reactor core'' that will guide the students, young nuclear engineering professionals, through designing, calculating, approval, and assembling a new nuclear reactor core. Students, young professionals from the South African nuclear industry, face the situation when a new nuclear reactor core is to be build from scratch. Several reactor core design options are pre-calculated. The selected design is re-calculated by the students, the result is then scrutinized by the regulator and, once all the analysis is approved, physical dismantling of the current core and assembling of the new core is done by the students, under a close supervision of the CTU staff. Finally the reactor is made critical with the new core. The presentation focuses on practical issues of such a course, desired reactor features and namely pedagogical and safety aspects. (orig.)

  10. Modelling activity transport behavior in PWR plant

    International Nuclear Information System (INIS)

    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)

  11. Current Status of Core and Advanced Adult Gastrointestinal Endoscopy Training in Canada: Survey of Existing Accredited Programs

    OpenAIRE

    Xiong, Xin; Barkun, Alan N; Waschke, Kevin; Martel, Myriam

    2013-01-01

    OBJECTIVE: To determine the current status of core and advanced adult gastroenterology training in Canada.METHODS: A survey consisting of 20 questions pertaining to core and advanced endoscopy training was circulated to 14 accredited adult gastroenterology residency program directors. For continuous variables, median and range were analyzed; for categorical variables, percentage and associated 95% CIs were analyzed.RESULTS: All 14 programs responded to the survey. The median number of core tr...

  12. Special Advanced Course for Core Sciences to Bring Up Project Leaders

    Science.gov (United States)

    Inagaki, Kenji; Tabata, Nobuhisa; Gofuku, Akio; Harada, Isao; Takada, Jun

    Special Advanced Course for Core Sciences has been introduced recently to Graduate School of Natural Science and Technology, Okayama University, to bring up a project leader. The following points are key education goals in this program : (1) knowledge of core sciences, (2) communication ability by using English, and (3) wide viewpoints for researches. In order to accomplish these goals, several lectures for core sciences, patent systems and engineering ethics as well as long term internships by the collaboration with some regional companies have been put in practice. In this paper, we describe the outline of the program, educational effects, and our experiences. Then, we discuss how effective the program is for bringing up an engineer or a scientist who can lead sciences and technologies of their domains. This paper also describes current activities of the program.

  13. THEHYCO-3DT: Thermal hydrodynamic code for the 3 dimensional transient calculation of advanced LMFBR core

    Energy Technology Data Exchange (ETDEWEB)

    Vitruk, S.G.; Korsun, A.S. [Moscow Engineering Physics Institute (Russian Federation); Ushakov, P.A. [Institute of Physics and Power Engineering, Obninsk (R)] [and others

    1995-09-01

    The multilevel mathematical model of neutron thermal hydrodynamic processes in a passive safety core without assemblies duct walls and appropriate computer code SKETCH, consisted of thermal hydrodynamic module THEHYCO-3DT and neutron one, are described. A new effective discretization technique for energy, momentum and mass conservation equations is applied in hexagonal - z geometry. The model adequacy and applicability are presented. The results of the calculations show that the model and the computer code could be used in conceptual design of advanced reactors.

  14. Development of Safety Related Issues and Essential Technologies for the Small and Medium Size PWR

    International Nuclear Information System (INIS)

    The objective of the study is to complement the safety related issues and their regulatory guides, deduced as a result of the review of the nuclear code and standards to being applied widely, then also is to find out and take concrete shape to the essential technologies for the small and medium size PWRs. The main activities performed in the study are listed below: - The review of the nuclear code and standards to being applied widely - The review of the safety related issues and their regulatory guides to be complemented · The review of the current developed state for advanced PWRs · The investigation of the disagreements or complements between the nuclear code and standards and the inherent characteristics for advanced PWRs - The essential technologies for the small and medium size PWRs · The essential technologies * The in-service inspection of reactor vessel assembly and steam generators * The development of comprehensive vibration assesment program * The feasibility of the test, monitoring, inspection and maintenance of components according to the extended cycle core · The publication of the ISI concept for an advanced PWR · The publication of the CVAP development for an advanced PWR · The preparation for an analysis input deck for the audit code - The evaluation for advanced PWRs to be corresponded to the nuclear code and standards

  15. Advanced computational methods for the assessment of reactor core behaviour during reactivity initiated accidents. Final report

    International Nuclear Information System (INIS)

    The document at hand serves as the final report for the reactor safety research project RS1183 ''Advanced Computational Methods for the Assessment of Reactor Core Behavior During Reactivity-Initiated Accidents''. The work performed in the framework of this project was dedicated to the development, validation and application of advanced computational methods for the simulation of transients and accidents of nuclear installations. These simulation tools describe in particular the behavior of the reactor core (with respect to neutronics, thermal-hydraulics and thermal mechanics) at a very high level of detail. The overall goal of this project was the deployment of a modern nuclear computational chain which provides, besides advanced 3D tools for coupled neutronics/ thermal-hydraulics full core calculations, also appropriate tools for the generation of multi-group cross sections and Monte Carlo models for the verification of the individual calculational steps. This computational chain shall primarily be deployed for light water reactors (LWR), but should beyond that also be applicable for innovative reactor concepts. Thus, validation on computational benchmarks and critical experiments was of paramount importance. Finally, appropriate methods for uncertainty and sensitivity analysis were to be integrated into the computational framework, in order to assess and quantify the uncertainties due to insufficient knowledge of data, as well as due to methodological aspects.

  16. Space-dependent dynamics of PWR

    International Nuclear Information System (INIS)

    The azimuthal dependent reactor dynamics coupled to thermohydraulics are studied by using the neutron-flux and coolant temperature signals measured at an actual PWR. The second azimuthal mode of neutron-flux fluctuation was found, and the coupling of the mode to thermohydraulics of the coolant was suggested. The coherent coolant flow in the reactor core seems to sustain this spatial oscillation mode. (authors)

  17. Hydraulic benchmark data for PWR mixing vane grid

    International Nuclear Information System (INIS)

    The purpose of the present study is to present new hydraulic benchmark data obtained for PWR rod bundles for the purpose of benchmarking Computational Fluid Dynamics (CFD) models of the rod bundle. The flow field in a PWR fuel assembly downstream of structural grids which have mixing vane grids attached is very complex due to the geometry of the subchannel and the high axial component of the velocity field relative to the secondary flows which are used to enhance the heat transfer performance of the rod bundle. Westinghouse has a CFD methodology to model PWR rod bundles that was developed with prior benchmark test data. As improvements in testing techniques have become available, further PWR rod bundle testing is being performed to obtain advanced data which has high spatial and temporal resolution. This paper presents the advanced testing and benchmark data that has been obtained by Westinghouse through collaboration with Texas A&M University. (author)

  18. 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)

    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 235U 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

  19. Advanced BWR core component designs and the implications for SFD analysis

    International Nuclear Information System (INIS)

    Prior to the DF-4 boiling water reactor (BWR) severe fuel damage (SFD) experiment conducted at the Sandia National Laboratories in 1986, no experimental data base existed for guidance in modeling core component behavior under postulated severe accident conditions in commercial BWRs. This paper will present the lessons learned from the DF-4 experiment (and subsequent German CORA BWR SFD tests) and the impact on core models in the current generation of SFD codes. The DF-4 and CORA BWR test assemblies were modeled on the core component designs circa 1985; that is, the 8 x 8 fuel assembly with two water rods and a cruciform control blade constructed of B4C-filled tubelets. Within the past ten years, the state-of-the-art with respect to BWR core component development has out-distanced the current SFD experimental data base and SFD code capabilities. For example, modern BWR control blade design includes hafnium at the tips and top of each control blade wing for longer blade operating lifetimes; also water rods have been replaced by larger water channels for better neutronics economy; and fuel assemblies now contain partial-length fuel rods, again for better neutronics economy. This paper will also discuss the implications of these advanced fuel assembly and core component designs on severe accident progression and on the current SFD code capabilities

  20. Sophysticated systems for analysing standard signals of a PWR NPP for diagnostic purposes

    International Nuclear Information System (INIS)

    An expert system is presented, which was designed for WWER type nuclear power plants (NPP) with 440 MWe PWR units. The input of the expert system includes the most important technological parameters of the core and of the primary and secondary loops. The expert system consists of the reactor noise diagnostics system (RNDS) and the on-line analysis system (VERONA). RNDS processes the AC components of measured signals. The application of RNDS advanced results in the following fields: registration of base line spectra; identification and localization of in core vibration, core barrel motion, propagating disturbances and the beginning of boiling; estimation of rector parameters; sensor diagnostics. VERONA processes the DC components. The following estimates are displayed: the total power production, the power generation in each fuel assembly and at ten elevations, the heat balance. (author)

  1. Design and Assessment Approach on Advanced SFR Safety with Emphasis on the Core Disruptive Accident Issue

    International Nuclear Information System (INIS)

    The safety of future sodium cooled fast reactors (SFRs) will be achieved at the same level as that achieved for future light water reactors (LWRs). The concept of defence in depth, as widely applied to the design of LWRs, will be applied to the safety design of advanced SFRs. Through the prevention, detection and control of accidents, core disruptive accidents (CDAs) will be excluded from design basis events. Considering that the SFR reactor core is not the most reactive configuration, unlike in LWRs, design measures to prevent CDAs and to mitigate the consequences of them are being considered as provisions for beyond design basis events. To meet future nuclear energy system safey goals effectively, advanced SFR designs should exploit passive safety features to increase safety margins and to enhance reliability, i.e. prevention and/or mitigation of CDAs. In particular, the safety approach needed to eliminate severe recriticality will be highly desirable, because with this approach, severe accidents in SFRs can be simply regarded as being similar to LWRs. In addition, it is easier to make full use of the excellent heat transport characteristics of sodium coolant in achieving in-vessel cooling and the retention of post-accident core debris. (author)

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

    International Nuclear Information System (INIS)

    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

  3. System comparative analysis of the most advanced pressured water reactors (PWR, WWER) and boiling water reactors (BWR) projects with the aim to choose the reactors for NPP construction in Kazakhstan

    International Nuclear Information System (INIS)

    Full text: The official decision on construction of a Nuclear Power Plant (NPP) in Kazakhstan has been accepted by the Kazakhstan government. The results on the choice of the power reactors projects of the NPP are given in the report. The choice has been carried out with the aim to develop recommendation on reactors of the NPP for construction in Kazakhstan. The choice of the reactors was based on the system comparative analysis of the most advanced power reactors projects using 15 criteria system of the nuclear, radiating and ecological safety and economic competitiveness. Following Pressurized Water Reactor (PWR, WWR) projects have been subjected to the system comparative analysis: 1) Large Sized Reactors (700 MW(el) and up): such as EPR, developed by Germany Siemens and France Framatome companies; CANDU-9, heavy-water reactor, developed by Atomic Energy of Canada Ltd (AECL); System 80+, developed by ABB Combustion Engineering company, USA; KNGR, Korean reactor of the next generation, developed by Korea Power Engineering Company, Inc.; APWR, Japanese advanced reactor, developed by Japan Atomic Power Company, Japan, Mitsubishi Heavy Industries, Japan and Westinghouse Electric Company, USA; WWER-1000 (V-392) - development by Atomenergoproect /Gydropress, Russian Federation; EP 1000, European passive reactor, development by Westinghouse, USA/Genesi, Italy. 2) Medium Sized Reactors (300 MWe - 700 MWe): AP-600, passive PWR, developed by the Westinghouse company; CANDU-6, heavy-water reactor, developed by Atomic Energy of Canada Ltd (AECL); An-tilde-600, passive PWR, developed by Nuclear Power Institute of China; WWER-640, Russian passive reactor, developed by 0KB ''Gidropress'' Experimental and Design Office, Russian Federation; MS-600, developed by Mitsubishi Company; KSNP-600, developed by Korea Power Engineering Company, Inc., South Korea. 3) Small Sized Reactors (a few MWe- 300 MWe): IRIS, reactor of IV generation, developed by the International Corporation of 13

  4. Evaluation of advanced two-phase flow instrumentation in SCTF Core-1

    International Nuclear Information System (INIS)

    In the Slab Core Test Facility (SCTF) Core-I, advanced two-phase flow instruments have been provided by the USNRC to measure the thermohydraulic behavior in the primary system including pressure vessel during the end of blowdown, refill and reflood phases of a postulated loss-of-coolant accident in a pressurized water reactor. The advanced instruments are turbine meters, drag disks, γ-densitometers, spool pieces, liquid level detectors (LLD), fluid distribution grids (FDG), impedance probes (flag, prong and string probes), film probes, and video optical probes. This report presents evaluated results of the data from these instruments. Some instruments are quantitatively evaluated by comparing with the data from the conventional instruments or the other advanced instruments. Main conclusions are as follows: (1) The spool pieces and the γ-densitometers work well and provide satisfactory results; (2) Some of the turbine meters, the impedance probes and the film probes give partially reasonable results, but still more improvements are required; (3) Most of the LLDs, the FDGs, the impedance probes, and the film probes do not work well due to a hard cable corrosion, and (4) The video optical probes give clear image of the flow pattern. (author)

  5. PWR fuel: experience and development

    International Nuclear Information System (INIS)

    The start-up of the large French nuclear program has rapidly led FRAGEMA to be one of the first PWR fuel suppliers. FRAGEMA is a joint subsidiary of two companies whose scopes of supply are fully complementary: FRAMATOME (NSSS vendor) and COGEMA (nuclear fuel cycle service supplier). At the center of these two activities FRAGEMA is in charge of designing and marketing fuel assemblies. Assistance is also offered to nuclear power plant operators in all fuel related fields by providing a wide range of services and a number of specialized components. Over the past years a statistical data base has been accumulated on fuel assembly behaviour under various operating conditions. At the same time extensive experimental programs have been, set up to develop advanced products to cope with utilities needs in the future. An overview of these two sides of our experience is presented in the following

  6. PWR standardization: The French experience

    International Nuclear Information System (INIS)

    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)

  7. 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

  8. Pu-breeding feasibility in PWR

    International Nuclear Information System (INIS)

    This study addresses the issue of alternative pathways for breeding plutonium in a 900 MWe three loop thermal pressurized water reactor (PWR), either fueled with uranium fuel (3.5% U-235) or with mixed fuel (20% MOX). During the operation of a nuclear reactor the in-core neutron flux and the ex-core neutron flux are monitored with flux detectors. At the places where those detectors operate, the guide thimbles and the vessel wall, respectively, the neutron flux can be used to irradiate material samples. This paper investigates whether it would be possible to produce plutonium by breeding it at the walls of a PWR vessel and/or in the guide thimbles. The neutron flux in the reactor and the corresponding multi-group spectra are estimated with Monte Carlo simulations for different positions at the vessel wall of a PWR operating with either UO2 or MOX. Then the irradiation of fresh uranium samples at the vessel wall and in the guide thimbles are calculated and the isotopic composition of the irradiated samples are determined. The minimum irradiation period and the necessary minimum amount of fresh uranium to breed different grades of plutonium are derived

  9. A study of the advancement of a reactor core design environment

    International Nuclear Information System (INIS)

    Full text: During the years from 2002 to 2004 a joint project has been performed by IFE, Halden and Yonden Engineering Corporation, Japan, to develop an advanced reactor core design environment based on a communication method for controlling a reactor core code system efficiently from PCs in a distributed network. The advanced reactor core design environment is realized by using Microsoft Visual Basic and communication software based on the IFE product SoftwareBus. The project has been carried out based on the fact that a computer-aided design system has been under development at Yonden Engineering Corporation in order to perform efficiently fuel replacement calculation by Yonden's reactor design code system. In this system, the structure is such that the physics calculation code system runs on UNIX workstations (in parallel) performing the calculations, while the Man-Machine Interface for controlling the calculation programs run on PCs in a distributed network. It has been emphasised to develop a reliable, flexible, adaptable and user-friendly system, which is easy to maintain. Therefore, a rather general communication tool (IFE's SoftwareBus) has been used for realizing communication of the n-pair n-node between the reactor core design code system and the PC applications. Further, a method of improvement in the speed of the optimal pattern calculation has been implemented by assigning each examination pattern to two or more computers distributed in the network and assigning the next pattern calculation to the computer, where the calculation has ended or has the lowest workload. The high-speed technology of the pattern survey by network distributed processing is based on SoftwareBus. The reactor core design code system is developed in FORTRAN running on a UNIX workstation (Solaris). The PC applications have been developed by using Microsoft Visual Basic on Windows 2000 platform. The first step of the verification and validation process was carried out in March

  10. Reactor physics analyses of the advanced neutron source three-element core

    International Nuclear Information System (INIS)

    A reactor physics analysis was performed for the Advanced Neutron Source reactor with a three-element core configuration. The analysis was performed with a two-dimensional r-z 20-energy-group finite-difference diffusion theory model of the 17-d fuel cycle. The model included equivalent r-z geometry representations of the central control rods, the irradiation and production targets, and reflector components. Calculated quantities include fuel cycle parameters, fuel element power distributions, unperturbed neutron fluxes in the reflector and target regions, reactivity perturbations, and neutron kinetics parameters

  11. A Preliminary Calculation of Annular Core Design for a High-flux Advanced Research Reactor

    International Nuclear Information System (INIS)

    Many of research reactors in operation over the world become old and the number of research reactors is expected to be reduced around 1/3 within a next decade. So it may be necessary to prepare in advance for the future demands of research reactors with a high performance. Therefore, based on the HANARO experiences through design to operation, a concept development of an improved research reactor is under doing. In this paper, 10 MW conceptual annular core is proposed and its basic characteristics were analyzed as a preliminary step

  12. Comparative sodium void effects for different advanced liquid metal reactor fuel and core designs

    International Nuclear Information System (INIS)

    An analysis of metal-, oxide-, and nitride-fueled advanced liquid metal reactor cores was performed to investigate the calculated differences in sodium void reactivity, and to determine the relationship between sodium void reactivity and burnup reactivity swing using the three fuel types. The results of this analysis indicate that nitride fuel has the least positive sodium void reactivity for any given burnup reactivity swing. Thus, it appears that a good design compromise between transient overpower and loss of flow response is obtained using nitride fuel. Additional studies were made to understand these and other nitride advantages. (author)

  13. Resource management in radio access and IP-based core networks for IMT Advanced and Beyond

    Institute of Scientific and Technical Information of China (English)

    SU Gang; HIDELL Markus; ABRAHAMSSON Henrik; AHLGREN Bengt; LI Dan; SJDIN Peter; TANYINGYONG Voravit; XU Ke

    2013-01-01

    The increased capacity needs, primarily driven by content distribution, and the vision of Internet-of-Things with billions of connected devices pose radically new demands on future wireless and mobile systems. In general the increased diversity and scale result in complex resource management and optimization problems in both radio access networks and the wired core network infrastructure. We summarize results in this area from a collaborative Sino-Swedish project within IMT Advanced and Beyond, covering adaptive radio resource management, energy-aware routing, OpenFlow-based network virtualization, data center networking, and access network caching for TV on demand.

  14. Progress of PWR reactor fuels: OSIRIS equipments

    International Nuclear Information System (INIS)

    The experimental reactor Osiris situated at the Saclay Nuclear Centre is a reactor fitted with tests and monitoring facilities. Of the pool and open core type, it can test the test fuel of PWR power stations under high neutron flux. The characteristic stresses of the operating states of power reactors can be reproduced in experimental devices suited to the various study subjects, be this the creep and deformation of zircaloy claddings, the behavior of fuel rods to power ramps, to load following, to remote regulation, to the cooling state in double phase or just analytical tests. The experimental irradiation devices extend from the single static coolant capsule, such as the NaK alloy, to the dynamic coolant test loop that operates in the cooling conditions representative of PWR's including water chemistry. Ancillary devices make it possible to carry out examinations and non-destructive testing: immersed neutron radiography, gamma scanning visualization monitoring device, eddy currents, profilometering

  15. 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

  16. 桩核修复新进展%Recent advance in post core restoration

    Institute of Scientific and Technical Information of China (English)

    周晓梅; 高兰敏

    2011-01-01

    根管治疗后的患牙大多有较严重的牙体缺损,容易发生冠根折裂.为了防止治疗失败,在修复时要考虑到无髓牙的薄弱易折性,给予适当的加强和弥补,良好的修复重建是十分必要的.桩核可以较好地恢复冠部缺损,为冠部固位提供支持力.根管治疗后的牙齿何时使用桩核,使用什么类型的桩核,许多临床工作者并不是十分清楚.纤维加强型树脂桩核具有一些理想的物理机械性能,它的出现冲击了桩核的一些经典理论,预示着很好的发展潜力.树脂水门汀粘接技术的迅猛发展也间接地推动了桩核的发展.本文总结以上几方面文献的研究结果,希望对口腔临床工作者全面了解桩核有所帮助,为临床工作提供依据.%After endodontic treatment, most affected teeth are seriously defecting, and are prone to crown root fracture. To avoid such treatment failure, the vulnerability and fragility of pulplessteeth should be taken into consideration by strengthening and compensating these teeth in restoration. Post core can well restore the defect of crown root and also provide support for its retention. However, clinicians are not very clear about when to use post core after endodontic treatment and what type of post core to use. The excellent physical and mechanical properties of fiber reinforced resin post core reformed the conventional theory on post core and suggest very promising future application. In addition, the rapid advance of resin cement adhesive technology also contributes indirectly to the development of post core. This article is a review of the literature of the aforementioned aspects as an effort to provide dental practitioners with comprehensive knowledge of post core and thus guide for clinical practice.

  17. SCORPIO-VVER core monitoring and surveillance system with advanced capabilities

    International Nuclear Information System (INIS)

    The SCORPIO-VVER system includes following features: 1) Validation of plant measurements and identification of sensor failures. 2) Optimum combination of measurements and calculations to obtain precise values of important parameters. 3) On-line 3D power distribution calculation with pin power reconstruction. 4) Limit checking and thermal margin calculation allowing for surveillance of VVER core limits such as DNBR, Sub-cooling margin, FdH and FQ peeking factors. 5) Integrated modules for monitoring fuel performance and coolant activity for identification of fuel failures. 6) Predictive capabilities and strategy planning, offering the possibility to check the consequences of operational manoeuvres in advance, prediction of critical parameters, etc. 7) Convenient monitoring of approach to criticality during reactor start-up. 8) Automated transition between cycles (fuel reload). The SCORPIO-VVER core monitoring system with its flexible and modular framework successfully responses to the plant operating needs and advances in nuclear fuel cycle strategies and fuel design. Modular framework allows for easy modifications of the system and implementation of new methods in physical modules. Even if the system is installed only on VVER-440 reactors, it could be adapted for VVER-1000 needs

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

    International Nuclear Information System (INIS)

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

  19. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2010

    Energy Technology Data Exchange (ETDEWEB)

    Rahmat Aryaeinejad; Douglas S. Crawford; Mark D. DeHart; George W. Griffith; D. Scott Lucas; Joseph W. Nielsen; David W. Nigg; James R. Parry; Jorge Navarro

    2010-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V&V, within the next 3-4 years via the ATR Core Modeling and Simulation and V&V Update (or “Core Modeling Update”) Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).

  20. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2010

    International Nuclear Information System (INIS)

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V and V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V and V, within the next 3-4 years via the ATR Core Modeling and Simulation and V and V Update (or 'Core Modeling Update') Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).

  1. Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

    International Nuclear Information System (INIS)

    The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

  2. Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Ragusa, Jean; Vierow, Karen

    2011-09-01

    The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

  3. PWR core response to boron dilution transient

    International Nuclear Information System (INIS)

    This paper illustrates the steps followed in order to set up a tool (composed of a plant model and of a procedure) that allows accounting for boron reactivity feedbacks during plant transients. The procedure that has been developed allows to find out the values of the boron feedback coefficients, given the differential boron worth, and to properly initialize the Thermal Hydraulic and the Neutronic (TH/NEU) system. Once the tool has been developed, it has been used to analyze different scenarios, resulting from deborated water injection from the reactor make-up system. The most important parameter, during this Reactivity Insertion Accidents (RIAs), is the Energy Released to the Fuel (ERF) and it has been monitored, in order to identify the situations when the fuel might be damaged (ERF > 250 kJ/kg, for high burnup fuel). The analyses have been performed using the RELAP5-3D computer code. The conclusion of the study is that the limited capability of modeling mixing phenomena provided by most common plant codes (such as RELAP5-3D) is not suitable to perform BE analyses of RIAs, since those accidents are so sensitive to boron concentration changes that the effect of uncertainties cannot be neglected. The use of Computational Fluid Dynamics (CFD) codes could reduce uncertainties enough to perform BE analyses and thus it should be recommended. (author)

  4. CORE

    DEFF Research Database (Denmark)

    Krigslund, Jeppe; Hansen, Jonas; Hundebøll, Martin;

    2013-01-01

    different flows. Instead of maintaining these approaches separate, we propose a protocol (CORE) that brings together these coding mechanisms. Our protocol uses random linear network coding (RLNC) for intra- session coding but allows nodes in the network to setup inter- session coding regions where flows...... increase the benefits of XORing by exploiting the underlying RLNC structure of individual flows. This goes beyond providing additional reliability to each individual session and beyond exploiting coding opportunistically. Our numerical results show that CORE outperforms both forwarding and COPE......-like schemes in general. More importantly, we show gains of up to 4 fold over COPE-like schemes in terms of transmissions per packet in one of the investigated topologies....

  5. Analysis of advanced sodium-cooled fast reactor core designs with improved safety characteristics

    International Nuclear Information System (INIS)

    Currently, the large majority of nuclear power plants are operated with thermal-neutron spectra and need regular fuel loading of enriched uranium. According to the identified conventional uranium resources and their current consumption rate, only about 100 years’ nuclear fuel supply is foreseen. A reactor operated with a fast-neutron spectrum, on the other hand, can induce self-sustaining, or even breeding, conditions for its inventory of fissile material, which effectively allow it, after the initial loading, to be refueled using simply natural or depleted uranium. This implies a much more efficient use of uranium resources. Moreover, minor actinides become fissionable in a fast-neutron spectrum, enabling full closure of the fuel cycle and leading to a minimization of long-lived radioactive wastes. The sodium-cooled fast reactor (SFR) is one of the most promising candidates to meet the Generation IV International Forum (GIF) declared goals. In comparison to other Generation IV systems, there is considerable design experience related to the SFR, and also more than 300 reactor years of practical operation. As a fast-neutron-spectrum system, the long-term operation of an SFR core in a closed fuel cycle will lead to an equilibrium state, where both reactivity and fuel mass flow stabilize. Although the SFR has many advantageous characteristics, it has one dominating neutronics drawback: there is generally a positive reactivity effect when sodium coolant is removed from the core. This so-called sodium void effect becomes even stronger in the equilibrium closed fuel cycle. The goal of the present doctoral research is to improve the safety characteristics of advanced SFR core designs, in particular, from the viewpoint of the positive sodium void reactivity effect. In this context, particular importance has been given to the dynamic core behavior under a hypothetical unprotected loss-of-flow (ULOF) accident scenario, in which sodium boiling occurs. The proposed

  6. Good practices in development of advanced assembly/core calculation methods and implementations of AEGIS/SCOPE2

    International Nuclear Information System (INIS)

    This paper reviews the history of development of AEGIS/SCOPE2, an advanced in-core fuel management code for PWRs. The initial project, development of a proto-type code, was started in 1996 as a feasibility study of the advanced calculation method/algorithm for advanced computation environments such as distributed parallel computers like PC-clusters which are commonly used nowadays. With success of development of the prototype code, a production-level advanced core calculation code, SCOPE2, was developed followed by AEGIS, an advanced assembly calculation code. These codes have been developed on the basis of the object-oriented programming approach and the agile software development. The authors extracted the key factors for success of the project as good practices from the viewpoint of code design, implementation, project management and verification and validation. Those practices are universal and may be applicable to any projects in the future. (author)

  7. A CFD Modeling Study for the Design of an Advanced HANARO Reactor Core Structure

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong-Hark; Chae, Hee-Teak; Park, Cheol; Kim, Heo-Nil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2007-07-01

    AHR(Advanced HANARO Reactor) based on HANARO has been under a conceptually designed with new ideas to implement new findings, which have been revealed from twelve years operation of HANARO. For example, a perforated structure to reduce the FIV(Flow Induced Vibration) of a fuel assembly has been considered to install. And a change of dual outlets to a single outlet has also been investigated to promote the accessibility and to work easily in the reactor pool. Those investigations have been conducted by the CFD (Computational Fluid Dynamics) method, which can provide us with an good understanding of three dimensional flow fields influenced by design changes without an experiment. In this study a CFD modeling study for an AHR core structure design is described.

  8. A CFD Modeling Study for the Design of an Advanced HANARO Reactor Core Structure

    International Nuclear Information System (INIS)

    AHR(Advanced HANARO Reactor) based on HANARO has been under a conceptually designed with new ideas to implement new findings, which have been revealed from twelve years operation of HANARO. For example, a perforated structure to reduce the FIV(Flow Induced Vibration) of a fuel assembly has been considered to install. And a change of dual outlets to a single outlet has also been investigated to promote the accessibility and to work easily in the reactor pool. Those investigations have been conducted by the CFD (Computational Fluid Dynamics) method, which can provide us with an good understanding of three dimensional flow fields influenced by design changes without an experiment. In this study a CFD modeling study for an AHR core structure design is described

  9. Evaluation of advanced cooling therapy's esophageal cooling device for core temperature control.

    Science.gov (United States)

    Naiman, Melissa; Shanley, Patrick; Garrett, Frank; Kulstad, Erik

    2016-05-01

    Managing core temperature is critical to patient outcomes in a wide range of clinical scenarios. Previous devices designed to perform temperature management required a trade-off between invasiveness and temperature modulation efficiency. The Esophageal Cooling Device, made by Advanced Cooling Therapy (Chicago, IL), was developed to optimize warming and cooling efficiency through an easy and low risk procedure that leverages heat transfer through convection and conduction. Clinical data from cardiac arrest, fever, and critical burn patients indicate that the Esophageal Cooling Device performs very well both in terms of temperature modulation (cooling rates of approximately 1.3°C/hour, warming of up to 0.5°C/hour) and maintaining temperature stability (variation around goal temperature ± 0.3°C). Physicians have reported that device performance is comparable to the performance of intravascular temperature management techniques and superior to the performance of surface devices, while avoiding the downsides associated with both. PMID:27043177

  10. Thermodynamic modelling of PWR coolant

    International Nuclear Information System (INIS)

    Corrosion products released from PWR and VVER primary circuit surface oxides are transported in the coolant to the core, where they deposit and are activated to form radioactive corrosion products, which can be re-released to re-deposit on out-of-core surfaces. Spinel solubilities vary with the pH, temperature and sometimes the hydrogen concentration of the coolant. This paper describes the development of an equilibrium thermodynamic model to predict such changes, and discusses the extent of the available solubility data for Fe, Ni, Co and Zn oxides. Results are described on the relative solubility of Fe and Ni under both normal operating conditions and during shutdown/start-up, and on the relative stabilities of stoichiometric and non-stoichiometric zinc ferrite spinels. Comparison of the calculated corrosion product concentrations with reactor measurements indicates that, in reactors with low Ni content in the steam generator alloys, the concentration of Ni in the coolant is limited by its availability in the surface oxide. In reactors with high-Ni alloys, the circulating Ni concentrations may be dominated by colloidal material. The calculated changes in Ni and Fe concentrations during the acid-reducing phase of shutdown are in reasonable agreement with measurements from Sizewell B. The paper highlights the need for a more comprehensive open corrosion product data base, the need to consider both boiling and radiolysis in the core on corrosion product solubility in different parts of the primary circuit and, finally, the importance of kinetic factors at low temperature behaviour during shutdown and start-up. (author)

  11. Maintenance robot for PWR plant

    International Nuclear Information System (INIS)

    The remote operation, automatic machines utilized in the field of the maintenance of component machinery and equipment in nuclear power plants, so-called maintenance robots, have produced effects in the reduction of radiation exposure, the improvement of the quality of working, the shortening of working time and so on, but still many robots have their specialized functions. The expectation of present day society to robots has been diversified, and the technical development of high function robots is advanced positively. In this report, the recent examples of the high function robots developed for PWR power stations with the support of technical progress and the trend of the technical development are explained. The needs and seeds of maintenance robot development are discussed. As the examples of heightening the functions of maintenance robots, the next generation ultrasonic testing machine highly advanced by sensor technology and size and weight reduction mechanism technology, the intelligent monitoring system for welding using AI technology and other manpower-saving robots are shown. (K.I.)

  12. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2012

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg, Principal Investigator; Kevin A. Steuhm, Project Manager

    2012-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to properly verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the next anticipated ATR Core Internals Changeout (CIC) in the 2014-2015 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its third full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (HELIOS, KENO6/SCALE, NEWT/SCALE, ATTILA, and an extended implementation of MCNP5) has been installed at the INL under various licensing arrangements. Corresponding models of the ATR and ATRC are now operational with all five codes, demonstrating the basic feasibility of the new code packages for their intended purpose. Of particular importance, a set of as-run core

  13. Simulation of small break loss of coolant accident in pressurized water reactor (PWR)

    International Nuclear Information System (INIS)

    A major safety concern in pressurized-water-reactor (PWR) design is the loss-of-coolant accident (LOCA),in which a break in the primary coolant circuit leads to depressurization, boiling of the coolant, consequent reduced cooling of the reactor core, and , unless remedial measures are taken, overheating of the fuel rods. This concern has led to the development of several simulators for safety analysis. This study demonstrates how the passive and active safety systems in conventional and advanced PWR behave during the small break loss of Coolant Accident (SBLOCA). The consequences of SBOLOCA have been simulated using IAEA Generic pressurized Water Reactor Simulator (GPWRS) and personal Computer Transient analyzer (PCTRAN) . The results were presented and discussed. The study has confirmed the major safety advantage of passive plants versus conventional PWRs is that the passive safety systems provide long-term core cooling and decay heat removal without the need for operator actions and without reliance on active safety-related system. (Author)

  14. Advanced in-core monitoring system for high-power reactors

    International Nuclear Information System (INIS)

    This paper encompasses such section as objective, conception and engineering solution for construction of advanced in-core instrumentation system for high power reactor, including WWER-1000. The ICIS main task is known to be an on-line monitoring of power distribution and functionals independently of design programs to avoid a common cause error. This paper shows in what way the recovery of power distribution has been carried out using the signals from in-core neutron detectors or temperature sensors. On the basis of both measured and processed data, the signals of preventive and emergency protection on local parameters (linear power of the maximum intensive fuel rods, departure from nucleate boiling ratio peaking factor) have been automatically generated. The paper presents a detection technology and processing methods for signals from SPNDs and TCs, ICIS composition and structure, computer hardware, system and applied software. Structure, composition and the taken decisions allow combining class IE and class B and C tasks in accordance with international standards of separation and safety category realization. Nowadays, ICIS-M is a system that is capable to ensure: monitoring, safety, information display and diagnostics function, which allow securing actual increase of quality, reliability and safety in operation of nuclear fuel and power units. Meanwhile, it reduce negative influence of human factor on thermal technical reliability in the operational process (Authors)

  15. Coupling of RMC and CFX for analysis of Pebble Bed-Advanced High Temperature Reactor core

    International Nuclear Information System (INIS)

    Highlights: ► The CFD code CFX is used for whole pebble bed reactor core calculation. ► The Monte Carlo Code RMC and CFX are used for the coupling of neutronics and T-H. ► Coupled calculations for steady-state problem can reach stable results. ► Increasing the number of neutron histories is effective to improve accuracy. - Abstract: This paper introduces a steady-state coupled calculation method using the Monte Carlo Code RMC (Reactor Monte Carlo) and the Computational Fluid Dynamic (CFD) code CFX for the analysis of a Pebble Bed-Advanced High Temperature Reactor (PB-AHTR) core. The RMC code is used for neutronics calculation while CFX is used for Thermal-Hydraulics (T-H) calculation. The porous media model is used in CFX modeling to simulate the pebble bed structure in PB-AHTR. The CFX model has also been validated against the RELAP5-3D model developed in the previous research. The script language PERL is used as a development tool to manipulate and control the entire coupled calculation. This research gives the conclusion that the steady-state coupled calculation using RMC and CFX is feasible and can obtain stable results within a few iterations. However, due to the statistical errors of Monte Carlo method, the fluctuation of results still occurs. For the purpose of improving the accuracy, the paper applies and discusses two methods, of which increasing the number of neutron histories is an effective method.

  16. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2013

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg

    2013-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for effective application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).

  17. Minimization of PWR reactor control rods wear

    International Nuclear Information System (INIS)

    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

  18. Stochastic optimization of loading pattern for PWR

    International Nuclear Information System (INIS)

    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

  19. PWR decontamination feasibility study

    International Nuclear Information System (INIS)

    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

  20. PWR type reactor

    International Nuclear Information System (INIS)

    From a PWR with a primary circuit, consisting of a reactor pressure vessel, a steam generator and a reactor coolant pump, hot coolant is removed by means of an auxiliary system containing h.p. pumps for feeding water into the primary circuit and being connected with a pipe, originating at the upper part, which has got at least one isolating value. This is done by opening an outlet in a part of the auxiliary system that has got a lower pressure than the reactor vessel. Preferably a water jet pump is used for mixing with the water of the auxiliary system. (orig.)

  1. Dismantling and decommissioning experience of commercial PWR

    International Nuclear Information System (INIS)

    Regarding the relatively youthness of FRAMATOME PWR's in operation none of these reactor needs to be decommissioned before 1992. However feasibility studies have been carried out by FRAMATOME for an on site entombment of active components and heavy equipments. In the past, partial dismantling of the reactor internals of the CHOOZ reactor: PWR of 320 MWe and a complete removal of the thermal shield protecting the reactor vessel were conducted successfully. After repair, the reactor power output has been upgraded of 10% and the reactor operates satisfactorily since 1970. More recently the discovery of scarce defects affecting centering pins of control guide tube located in the upper reactor internals of 900 MWe plants has initiated the construction of several ''Hot stand equipments'' for the systematic replacement of these centering pins. FRAMATOME is presently actively studying possible options consisting either to extend the plant life beyond its initial licence life, or to convert classical PWR into an advanced reactor more economical in terms of uranium consumption

  2. Evolution of reactor monitoring and protection systems for PWR

    International Nuclear Information System (INIS)

    This paper presents the evolution of the reactor protection systems and of the reactor monitoring systems for PWR since the initial design in the Fessenheim plant to the latest development for the EPR (European pressurized reactor). The features of both systems for the different kinds of PWR operating in France: 900 MWe, 1300 MWe and N4, are reviewed. The expected development of powerful micro-processors for computation, for data analysis and data storage will make possible in a near future the monitoring on a 3-dimensional basis and on a continuous manner, of the nuclear power released in the core. (A.C.)

  3. Corrosion product transfer in PWR primary circuits during cold shutdowns

    International Nuclear Information System (INIS)

    Two experimental tests have been performed to study the corrosion product transfer during PWR cold shutdowns: one with nickel ferrite and the other one with metallic nickel. The temperature evolution together with boron and oxygen concentration evolutions are similar to those obtained during cold shutdowns of PWR primary circuits. With metallic nickel, the increase of the Ni concentration occurs during the decrease of the primary temperature and mainly when the oxygenation is realised. Whereas, with nickel ferrite, the Ni concentration increase occurs during the 24 hours after the oxygenation. These results compared to the plant data lead to conclude that metallic nickel presence in the core is the most probable hypothesis. (author)

  4. Evaluation of PWR and BWR pin cell benchmark results

    International Nuclear Information System (INIS)

    In order to carry out reliable reactor core calculations for a boiled water reactor (BWR) or a pressurized water reactor (PWR) first reactivity calculations have to be carried out for which several calculation programs are available. The purpose of the title project is to exchange experiences to improve the knowledge of this reactivity calculations. In a large number of institutes reactivity calculations of PWR and BWR pin cells were executed by means of available computer codes. Results are compared. It is concluded that the variations in the calculated results are problem dependent. Part of the results is satisfactory. However, further research is necessary

  5. PHEDRE model for the simulation of PWR reactors

    International Nuclear Information System (INIS)

    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

  6. Application of Westinghouse NEXUS/ANC9 cross-section model for PWR accident analyses

    International Nuclear Information System (INIS)

    NEXUS/ANC9 is the latest licensed PWR core design code system developed by Westinghouse. This system has demonstrated capabilities of modeling advanced core designs with improved accuracy in core reactivity and power distribution predictions. NEXUS/ANC9 system is being rolled out to replace the current APA system (ALPHA/PHOENIX-P/ANC) for routine core calculations. In addition to the standard core design calculations, investigations are underway to explore the possibility to expand the NEXUS/ANC9 application for safety analysis, especially at accident conditions. The main focus of the investigation is the evaluation of the NEXUS/ANC9 cross-section representation model conditions like high void and significant change of core pressure. Comparisons of the predicted parameters among ANC9, PARAGON lattice code and MCNP calculations are presented. The results show that NEXUS/ANC9 is able to model the cross-section behavior and accurately reproduce lattice code results at all simulated conditions. (author)

  7. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2011

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Devin A. Steuhm

    2011-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V&V, within the next 3-4 years via the ATR Core Modeling and Simulation and V&V Update (or 'Core Modeling Update') Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the anticipated ATR Core Internals Changeout (CIC) in the 2014 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its first full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (SCALE, KENO-6, HELIOS, NEWT, and ATTILA) have been installed at the INL under various permanent sitewide license agreements and corresponding baseline models of the ATR and ATRC are now operational, demonstrating the basic feasibility of these code packages for their intended purpose. Furthermore

  8. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2011

    International Nuclear Information System (INIS)

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V and V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V and V, within the next 3-4 years via the ATR Core Modeling and Simulation and V and V Update (or 'Core Modeling Update') Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the anticipated ATR Core Internals Changeout (CIC) in the 2014 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its first full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (SCALE, KENO-6, HELIOS, NEWT, and ATTILA) have been installed at the INL under various permanent sitewide license agreements and corresponding baseline models of the ATR and ATRC are now operational, demonstrating the basic feasibility of these code packages for their intended purpose

  9. N4 PWR makes full use of distributed processing and local networks

    Energy Technology Data Exchange (ETDEWEB)

    Aschenbrenner, J.F.; Tetreau, F.; Colling, J.M.

    1988-01-01

    The new instrumentation and control systems for the French N4 PWR power plant make extensive use of programmable controllers based on advanced microprocessor technology and distributed processing. Local networking techniques are widely used which simplify architecture and equipment design.

  10. Molten core concrete interaction and development of core catcher. (2) Thermal shock effects for advanced high temperature ceramics

    International Nuclear Information System (INIS)

    Four monolithic refractory ceramics (Al2O3, SiC, TiC and TiN) were exposed to severe thermal shock caused by thermite reaction. Such experiments were intended to simulate the exposure of a core catcher to hot debris of melted core in nuclear reactor. Moreover, the chemical interactions between tested materials and products of thermite reaction were investigated by holding samples in laboratory furnace for 5 hours at 1000degC in air. None of the materials entirely withstand such sudden temperature rise, of approximately 1000degC/s, especially the sample made of Al2O3 peeled into small pieces under thermite mixture. Both TiN and TiC cracked in the central part, perpendicularly to the reaction front and proceeding heat wave. The best performance was observed in SiC sample, which is caused by the lowest thermal expansion coefficient. Considering the application of SiC in the construction of core catcher, research on thermal shock resistance improvements have to be performed. (author)

  11. Investigation, experiment and analysis on PWR sump screen clogging issue

    International Nuclear Information System (INIS)

    JNES has been conducting experimental and analytical study to develop an evaluation method concerning the downstream effect of the sump screen clogging issue during LOCA in PWR plants. Flow clogging characteristics were investigated based on data for the relation of pressure loss and flow velocity during flow clogging due to debris accumulation. Deposition of chemical precipitates on the fuel cladding using an electrically heated rod was investigated. A test shows chemical precipitates deposited on the cladding and the deposit was mainly analyzed to be calcium compounds. The analysis with a thermal-hydraulic code on the downstream effect has shown that the core could be cooled because the core inlet flow compensates a evaporation of coolant due to the decay-heat even if core inlet was 99% clogged just after the ECCS recirculation operation started during the cold-leg break LOCA in PWR plants. (author)

  12. Investigation, experiment and analysis on PWR sump screen clogging issue

    International Nuclear Information System (INIS)

    JNES has been conducting experimental and analytical study to develop an evaluation method concerning the chemical effect and the downstream effect of the sump screen clogging issue during LOCA in PWR plants. Chemical effect tests show that corrosion of carbon steel and galvanized steal may come to be important in domestic plants, in addition to corrosion of aluminum and insulator which has been considered dominant in the chemical effect. With respect to the downstream effect, deposition of chemical precipitates on the fuel cladding using an electrically heated rod is investigated. A test shows chemical precipitates deposited on the cladding and the deposit was mainly analyzed to be calcium compounds. The analysis on the downstream effect has shown that even if core inlet was completely clogged just after the recirculation operation started during LOCA in PWR plants, although upper part of core may be uncovered temporary and cladding temperature increased, core could be cooled by coolant injection through the hot-leg. (author)

  13. Assessment of subcriticality during PWR-type reactor refueling

    International Nuclear Information System (INIS)

    During the core loading period of a PWR, any fuel assembly misplacements may significantly reduce the existing criticality margin. The Dampierre 4-18 event showed the present monitoring based on the variations of the outside-core detector counting rate cannot detect such misplacements. In order to circumvent that, a more detailed analysis of the available signal was done. We particularly focused on the neutronic noise analysis methods such as MSM (modified source multiplication), MSA (amplified source multiplication), Rossi-α and Feynman-α methods. The experimental part of our work was dedicated to the application of those methods to a research reactor. Finally, our results showed that those methods cannot be used with the present PWR instrumentation. Various detector positions were then studied using Monte Carlo calculations capable of following the neutron origin. Our results showed that the present technology does not allow us to use any solution based on neutron detection for monitoring core loading. (author)

  14. Development of a standard data base for FBR core nuclear design. 7. Advances in JUPITER experiment analyses

    International Nuclear Information System (INIS)

    The present report compiles the advances in experiment analyses of JUPITER, which was joint research programs between U.S.DOE and PNC of Japan, using the Zero Power Physics Reactor (ZPPR) large fast critical facility at ANL-Idaho in 1978 to 1988. The advances here are use of the latest nuclear data library and the application of analytical methods which treat mechanisms in more detail or use fewer modeling approximations. As a result of using the latest nuclear data library, C/E values of nearly all characteristics approached unity, and the discrepancies between cores were reduced. Thus it is shown that the latest data library is effective for an analysis of nuclear characteristics. Further, an advance in analytical methods brought C/E value close to unity, and it clarifies the causes of differences between the calculational and experimental values. It is judged that the JUPITER integral data are very effective for the production of the unified nuclear constants which are intended for the core design of the demonstration fast breeder reactor. Furthermore, for improved accuracy in the analytical system, an advance in analytical methods for the evaluation of flux distributions in blanket regions is very effective. Especially, the accuracy for radially heterogeneous cores would be greatly improved by such an advance. (J.P.N.). 146 refs

  15. Preliminary fracture analysis of the core pressure boundary tube for the Advanced Neutron Source Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schulz, K.C. [Univ. of Turabo, Gurabo, Puerto (Puerto Rico). College of Engineering; Yahr, G.T. [Oak Ridge National Lab., TN (United States)

    1995-08-01

    The outer core pressure boundary tube (CPBT) of the Advanced neutron Source (ANS) reactor being designed at Oak Ridge National Laboratory is currently specified as being composed of 6061-T6 aluminum. ASME Boiler and Pressure Vessel Code fracture analysis rules for nuclear components are based on the use of ferritic steels; the expressions, tables, charts and equations were all developed from tests and analyses conducted for ferritic steels. Because of the nature of the Code, design with thin aluminum requires analytical approaches that do not directly follow the Code. The intent of this report is to present a methodology comparable to the ASME Code for ensuring the prevention of nonductile fracture of the CPBT in the ANS reactor. 6061-T6 aluminum is known to be a relatively brittle material; the linear elastic fracture mechanics (LEFM) approach is utilized to determine allowable flaw sizes for the CPBT. A J-analysis following the procedure developed by the Electric Power Research Institute was conducted as a check; the results matched those for the LEFM analysis for the cases analyzed. Since 6061-T6 is known to embrittle when irradiated, the reduction in K{sub Q} due to irradiation is considered in the analysis. In anticipation of probable requirements regarding maximum allowable flaw size, a survey of nondestructive inspection capabilities is also presented. A discussion of probabilistic fracture mechanics approaches, principally Monte Carlo techniques, is included in this report as an introduction to what quantifying the probability of nonductile failure of the CPBT may entail.

  16. Core library for advanced scenario simulation, C. L. A. S. S.: Principle and application

    International Nuclear Information System (INIS)

    The global warming, the increase of world population and the depletion of fossil resources have lead us in a major energy crisis. Using electronuclear energy could be one of the means to solve a part of these issues. The way out of this crisis may be enlightened by the study of transitional scenarios, guiding the political decisions. The reliability of those studies passes through the wide variety of the simulation tools and the comparison between them. From this perspective and in order to perform complex electronuclear scenario simulation, the open source Core Library for Advance Scenario Simulation (CLASS) is being developed. CLASS main asset is its ability to include any kind of reactor, whether the system is innovative or standard. A reactor is fully described by its evolution database that must contain a set of different fuel compositions in order to simulate transitional scenarios. CLASS aims at being a useful tool to study scenarios involving Generation IV reactors as well as innovative fuel cycles, like the Thorium cycle. The following contribution will present in detail the CLASS software. Starting with the working principle of this tool, one will explain the working process of the different modules such as the evolution module. It will be followed by an exhaustive presentation of the UOX-MOX bases generation procedure. Finally a brief analysis of the error made by the CLASS evolution module will be presented. (author)

  17. Chemical and physical analysis of core materials for advanced high temperature reactors with process heat applications

    International Nuclear Information System (INIS)

    Various chemical and physical methods for the analysis of structural materials have been developed in the research programmes for advanced high temperature reactors. These methods are discussed using as examples the structural materials of the reactor core - the fuel elements consisting of coated particles in a graphite matrix and the structural graphite. Emphasis is given to the methods of chemical analysis. The composition of fuel kernels is investigated using chemical analysis methods to determine the heavy metals content (uranium, plutonium, thorium and metallic impurity elements) and the amount of non-metallic constituents. The properties of the pyrocarbon and silicon carbide coatings of fuel elements are investigated using specially developed physiochemical methods. Regarding the irradiation behaviour of coated particles and fuel elements, methods have been developed for examining specimens in hot cells following exposures under reactor operating conditions, to supplement the measurements of in-reactor performance. For the structural graphite, the determination of impurities is important because certain impurities may cause pitting corrosion during irradiation. The localized analysis of very low impurity concentrations is carried out using spectrochemical d.c. arc excitation, local laser and inductively coupled plasma methods. (orig.)

  18. Preliminary fracture analysis of the core pressure boundary tube for the Advanced Neutron Source Research Reactor

    International Nuclear Information System (INIS)

    The outer core pressure boundary tube (CPBT) of the Advanced neutron Source (ANS) reactor being designed at Oak Ridge National Laboratory is currently specified as being composed of 6061-T6 aluminum. ASME Boiler and Pressure Vessel Code fracture analysis rules for nuclear components are based on the use of ferritic steels; the expressions, tables, charts and equations were all developed from tests and analyses conducted for ferritic steels. Because of the nature of the Code, design with thin aluminum requires analytical approaches that do not directly follow the Code. The intent of this report is to present a methodology comparable to the ASME Code for ensuring the prevention of nonductile fracture of the CPBT in the ANS reactor. 6061-T6 aluminum is known to be a relatively brittle material; the linear elastic fracture mechanics (LEFM) approach is utilized to determine allowable flaw sizes for the CPBT. A J-analysis following the procedure developed by the Electric Power Research Institute was conducted as a check; the results matched those for the LEFM analysis for the cases analyzed. Since 6061-T6 is known to embrittle when irradiated, the reduction in KQ due to irradiation is considered in the analysis. In anticipation of probable requirements regarding maximum allowable flaw size, a survey of nondestructive inspection capabilities is also presented. A discussion of probabilistic fracture mechanics approaches, principally Monte Carlo techniques, is included in this report as an introduction to what quantifying the probability of nonductile failure of the CPBT may entail

  19. The development of flow test technology for PWR fuel assemblies

    International Nuclear Information System (INIS)

    The objective of this project is to design and construct a high temperature and pressure flow test facility and to develop flow test technology for the evaluation of PWR fuel performance. For the nuclear fuel safety aspect it is of importance to evaluate the thermalhydraulic compatibility and mechanical integrity of a newly designed fuel through the design verification test. The PWR-Hot Test Loop facility is under construction to be used to perform a pressure drop test, a lift force test and a fretting corrosion test of a fullsize PWR fuel assembly at reactor operating conditions. This facility was designed to be used to produce the hydraulic parameters of the existing PWR fuel assemblies(14x14FA, 16x16FA, 17x17FA) and to verify a design of advanced fuel assemblies (KAFA-I and KAFA-II) developed by KAERI. The PWR-Cold Test Loop facility with the 5x5 Rod Bundles in the test section was designed and installed to carry out the flow distribution study by means of Laser Doppler Velocimeter. The LDV techniques have been developed and used to measure the flow velocity and turbulent intensity for evaluating mixing effects of a newly designed spacer grid with and without mixing vanes, cross flow between the fuel assemblies and a turbulent model. (Author)

  20. TRAC-P1: an advanced best estimate computer program for PWR LOCA analysis. I. Methods, models, user information, and programming details

    International Nuclear Information System (INIS)

    The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos Scientific Laboratory (LASL) to provide an advanced ''best estimate'' predictive capability for the analysis of postulated accidents in light water reactors (LWRs). TRAC-Pl provides this analysis capability for pressurized water reactors (PWRs) and for a wide variety of thermal-hydraulic experimental facilities. It features a three-dimensional treatment of the pressure vessel and associated internals; two-phase nonequilibrium hydrodynamics models; flow-regime-dependent constitutive equation treatment; reflood tracking capability for both bottom flood and falling film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The TRAC-Pl User's Manual is composed of two separate volumes. Volume I gives a description of the thermal-hydraulic models and numerical solution methods used in the code. Detailed programming and user information is also provided. Volume II presents the results of the developmental verification calculations

  1. In- and ex-vessel coupled analysis of IVR-ERVC phenomenon for large scale PWR

    International Nuclear Information System (INIS)

    Highlights: • MELCOR models are built for large scale PWR with thermal power reaching 5000 MWt. • In- and ex-vessel coupled transient analysis of IVR-ERVC phenomenon is performed. • Results show that the IVR-ERVC strategy is an effective way to maintain RPV integrity during a severe accident. - Abstract: As a key severe accident management strategy for light water reactors (LWRs), in-vessel retention (IVR) through external reactor vessel cooling (ERVC) has been the focus of relevant studies for decades. However, previous studies only investigated the molten pool configurations considered to be in a final steady state mainly for reactors of such as AP600 and AP1000. Furthermore, most of studies performed in the past dealt with analysis for an isolated IVR-ERVC process, without considering the strong coupling between the internal and external reactor pressure vessel (RPV) conditions. This paper addresses the IVR-ERVC issues from a transient perspective using the severe accident code MELCOR for a large advanced passive power plant: a three-loop, 5000 MWt scale pressurized water reactor with passive safety features. The analysis is mainly focused on the severe accident transients including core degradation and relocation, molten pool formation and growth, and heat transfer within a molten pool. Furthermore, internal and external RPV conditions are combined together in the IVR-ERVC analysis. MELCOR calculations for lower head heat flux are then compared with critical heat flux (CHF) to assess the effectiveness of IVR-ERVC. The results suggest that lower head heat flux is below the CHF value. Therefore, the IVR-ERVC strategy for this large PWR is considered to be feasible. It was also found that as the reactor power is raised to large scale PWR, new accident sequences may occur during the severe accident evolution, thus leading to a proposal of a completely new molten pool configuration for future studies

  2. TRAC-P1: an advanced best estimate computer program for PWR LOCA analysis. I. Methods, models, user information, and programming details

    Energy Technology Data Exchange (ETDEWEB)

    1978-05-01

    The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos Scientific Laboratory (LASL) to provide an advanced ''best estimate'' predictive capability for the analysis of postulated accidents in light water reactors (LWRs). TRAC-Pl provides this analysis capability for pressurized water reactors (PWRs) and for a wide variety of thermal-hydraulic experimental facilities. It features a three-dimensional treatment of the pressure vessel and associated internals; two-phase nonequilibrium hydrodynamics models; flow-regime-dependent constitutive equation treatment; reflood tracking capability for both bottom flood and falling film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The TRAC-Pl User's Manual is composed of two separate volumes. Volume I gives a description of the thermal-hydraulic models and numerical solution methods used in the code. Detailed programming and user information is also provided. Volume II presents the results of the developmental verification calculations.

  3. Validation of STAR-CCM+ for bouyancy driven mixing in a PWR reactor pressure vessel

    International Nuclear Information System (INIS)

    Within the OECD/NEA PKL-II project, experiments have been carried out aimed at investigating the flow mixing in the downcomer and lower plenum of a pressurized water reactor (PWR) in the buoyancy driven mixing regimes. The experiments have been performed at the ROCOM test facility, a 1:5 scaled representation of a KONVOI type pressurized water reactor (PWR). The facility is equipped with advanced instrumentation (i.e. wiremesh sensors) allowing a detailed measurement of flow mixing in the downcomer annulus and at the core inlet. A computational fluid dynamic (CFD) model has been developed at the Paul Scherrer Institute within the STARS project, employing the STAR-CCM+ code. The CFD model has been validated against the ROCOM experimental results. It has been shown that the developed model provided a good agreement with experiment. In order to evaluate the difference between momentum driven and density driven mixing regimes, calculations were performed assuming no density difference, and with 12% higher density in one of the loops respectively. (author)

  4. Validation of the new code package APOLLO2.8 for accurate PWR neutronics calculations

    International Nuclear Information System (INIS)

    This paper summarizes the Qualification work performed to demonstrate the accuracy of the new APOLLO2.S/SHEM-MOC package based on JEFF3.1.1 nuclear data file for the prediction of PWR neutronics parameters. This experimental validation is based on PWR mock-up critical experiments performed in the EOLE/MINERVE zero-power reactors and on P.I. Es on spent fuel assemblies from the French PWRs. The Calculation-Experiment comparison for the main design parameters is presented: reactivity of UOX and MOX lattices, depletion calculation and fuel inventory, reactivity loss with burnup, pin-by-pin power maps, Doppler coefficient, Moderator Temperature Coefficient, Void coefficient, UO2-Gd2O3 poisoning worth, Efficiency of Ag-In-Cd and B4C control rods, Reflector Saving for both standard 2-cm baffle and GEN3 advanced thick SS reflector. From this qualification process, calculation biases and associated uncertainties are derived. This code package APOLLO2.8 is already implemented in the ARCADIA new AREVA calculation chain for core physics and is currently under implementation in the future neutronics package of the French utility Electricite de France. (authors)

  5. Design and assessment approach on advanced SFR safety with emphasis on core disruptive accident issue

    International Nuclear Information System (INIS)

    evaluation is Beyond Design Basis Events with best-estimate method and assumptions. The purpose of CDA analysis has been therefore to provide or confirm an additional safety margin of the plant strictly designed for Design Basis Events. Generation IV Nuclear Energy Systems are being developed under the initiative of Generation IV International Forum (GIF) begun in 2000. The SFR was selected as one of the promising concepts together with other five concepts. Three goals for the Generation IV nuclear systems have been defined in the safety and reliability as listed below. - Safety and Reliability - 1, Generation IV nuclear energy systems operations will excel in safety and reliability. - Safety and Reliability - 2, Generation IV nuclear energy systems will have a very low likelihood and degree of reactor core damage. - Safety and Reliability - 3, Generation IV nuclear energy systems will eliminate the need for offsite emergency response. From a viewpoint of DiD philosophy, for the purpose of eliminating the need for the fifth level, which is the off-site emergency response, we need to strengthen the safety design of the fourth level of DiD, which is severe accident management. On the other hand, there is the fact that emergency response plans have been already prepared in compliance with national laws and regulations in many countries. In this sense it is effective to provide design measures to mitigate postulated severe accidents within a plant and/or to provide sufficient grace period to reach core damage and/or containment failure for the recovery by operator and for the judgement of proclamation of emergency response by authority taking into account the characteristic of severe accident progression. To effectively meet the Generation-IV systems goals, advanced SFR designs exploit passive safety features to increase safety margins and to enhance reliability. The system behavior will vary depending on system size, design features, and fuel type. R and D for passive safety

  6. 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)

    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

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

    International Nuclear Information System (INIS)

    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

  8. Application of PWR LOCA margin with the revised appendix K rule

    International Nuclear Information System (INIS)

    Today's focus for nuclear power plant utility owners is to improve plant performances such that the cost per kilowatthour is minimized with enhanced safety. This paper will discuss the impact of design and licensed margin on PWR plant performance, how these margins can be used to improve PWR performance, and how Westinghouse is addressing the regulatory design limits for large break and small break LOCA which impact core thermal design margin. (orig./GL)

  9. Initial global 2-D shielding analysis for the Advanced Neutron Source core and reflector

    Energy Technology Data Exchange (ETDEWEB)

    Bucholz, J.A.

    1995-08-01

    This document describes the initial global 2-D shielding analyses for the Advanced Neutron Source (ANS) reactor, the D{sub 2}O reflector, the reflector vessel, and the first 200 mm of light water beyond the reflector vessel. Flux files generated here will later serve as source terms in subsequent shielding analyses. In addition to reporting fluxes and other data at key points of interest, a major objective of this report was to document how these analyses were performed, the phenomena that were included, and checks that were made to verify that these phenomena were properly modeled. In these shielding analyses, the fixed neutron source distribution in the core was based on the `lifetime-averaged` spatial power distribution. Secondary gamma production cross sections in the fuel were modified so as to account intrinsically for delayed fission gammas in the fuel as well as prompt fission gammas. In and near the fuel, this increased the low-energy gamma fluxes by 50 to 250%, but out near the reflector vessel, these same fluxes changed by only a few percent. Sensitivity studies with respect to mesh size were performed, and a new 2-D mesh distribution developed after some problems were discovered with respect to the use of numerous elongated mesh cells in the reflector. All of the shielding analyses were performed sing the ANSL-V 39n/44g coupled library with 25 thermal neutron groups in order to obtain a rigorous representation of the thermal neutron spectrum throughout the reflector. Because of upscatter in the heavy water, convergence was very slow. Ultimately, the fission cross section in the various materials had to be artificially modified in order to solve this fixed source problem as an eigenvalue problem and invoke the Vondy error-mode extrapolation technique which greatly accelerated convergence in the large 2-D RZ DORT analyses. While this was quite effective, 150 outer iterations (over energy) were still required.

  10. Initial global 2-D shielding analysis for the Advanced Neutron Source core and reflector

    International Nuclear Information System (INIS)

    This document describes the initial global 2-D shielding analyses for the Advanced Neutron Source (ANS) reactor, the D2O reflector, the reflector vessel, and the first 200 mm of light water beyond the reflector vessel. Flux files generated here will later serve as source terms in subsequent shielding analyses. In addition to reporting fluxes and other data at key points of interest, a major objective of this report was to document how these analyses were performed, the phenomena that were included, and checks that were made to verify that these phenomena were properly modeled. In these shielding analyses, the fixed neutron source distribution in the core was based on the 'lifetime-averaged' spatial power distribution. Secondary gamma production cross sections in the fuel were modified so as to account intrinsically for delayed fission gammas in the fuel as well as prompt fission gammas. In and near the fuel, this increased the low-energy gamma fluxes by 50 to 250%, but out near the reflector vessel, these same fluxes changed by only a few percent. Sensitivity studies with respect to mesh size were performed, and a new 2-D mesh distribution developed after some problems were discovered with respect to the use of numerous elongated mesh cells in the reflector. All of the shielding analyses were performed sing the ANSL-V 39n/44g coupled library with 25 thermal neutron groups in order to obtain a rigorous representation of the thermal neutron spectrum throughout the reflector. Because of upscatter in the heavy water, convergence was very slow. Ultimately, the fission cross section in the various materials had to be artificially modified in order to solve this fixed source problem as an eigenvalue problem and invoke the Vondy error-mode extrapolation technique which greatly accelerated convergence in the large 2-D RZ DORT analyses. While this was quite effective, 150 outer iterations (over energy) were still required

  11. AREVA NP's advanced Thermal Hydraulic Methods for Reactor Core and Fuel Assembly Design

    International Nuclear Information System (INIS)

    AREVA NP, two converged sub-channel codes have been defined: the homogenous equilibrium model (HEM) code COBRA-FLX and the multi fluid field code F-COBRA-TF. Apart from the sub-channel codes and some smaller specialized codes computational fluid dynamic (CFD) codes are the second important pillar of the AREVA TH code strategy. In the last decade big improvements in the available codes were made and the computing power increased dramatically. Consequently CFD became a reliable and robust tool; thanks to the increased computing power the size of the efficiently calculable models became large enough to be interesting for TH application in fuel assemblies. The main potential of CFD originates from the fact that CFD can predict TH quantities directly, based on the geometric information stored in a computer aided design (CAD) file for mechanic design, the tabulated fluid properties and the desired operating parameters. Hence CFD can be seen as a tool which can be used to perform virtual TH experiments. But unlike experiments where often the access is limited to few TH quantities, CFD provides the comprehensive local TH information and valuable insight into length scales smaller than sub-channels cross sections. Thus, CFD cannot only be used to directly determine the interesting quantities, but also to complement experiments and sub-channel code analysis as well as to support further development of sub-channel codes. AREVA NP's TH methods and codes development strategy follows thus two main streams: 1. Updating and improving the sub-channel codes in order to meet the advanced customer and licensing requirements like improved physical modeling, more detailed information, more flexibility, etc. The recent developments cover the following domains: a. Improved/ Advanced Physics; b. Improved Coding/ Advanced Algorithm. Objective: faster code allowing to perform more calculations or to calculate large models (Pin-by-Pin full core calculations steady state and transient); c

  12. High accuracy modeling for advanced nuclear reactor core designs using Monte Carlo based coupled calculations

    Science.gov (United States)

    Espel, Federico Puente

    The main objective of this PhD research is to develop a high accuracy modeling tool using a Monte Carlo based coupled system. The presented research comprises the development of models to include the thermal-hydraulic feedback to the Monte Carlo method and speed-up mechanisms to accelerate the Monte Carlo criticality calculation. Presently, deterministic codes based on the diffusion approximation of the Boltzmann transport equation, coupled with channel-based (or sub-channel based) thermal-hydraulic codes, carry out the three-dimensional (3-D) reactor core calculations of the Light Water Reactors (LWRs). These deterministic codes utilize nuclear homogenized data (normally over large spatial zones, consisting of fuel assembly or parts of fuel assembly, and in the best case, over small spatial zones, consisting of pin cell), which is functionalized in terms of thermal-hydraulic feedback parameters (in the form of off-line pre-generated cross-section libraries). High accuracy modeling is required for advanced nuclear reactor core designs that present increased geometry complexity and material heterogeneity. Such high-fidelity methods take advantage of the recent progress in computation technology and coupled neutron transport solutions with thermal-hydraulic feedback models on pin or even on sub-pin level (in terms of spatial scale). The continuous energy Monte Carlo method is well suited for solving such core environments with the detailed representation of the complicated 3-D problem. The major advantages of the Monte Carlo method over the deterministic methods are the continuous energy treatment and the exact 3-D geometry modeling. However, the Monte Carlo method involves vast computational time. The interest in Monte Carlo methods has increased thanks to the improvements of the capabilities of high performance computers. Coupled Monte-Carlo calculations can serve as reference solutions for verifying high-fidelity coupled deterministic neutron transport methods

  13. Qualification according to PDI's techniques UT EPRI methodology Phased Array for the inspection of vessels of PWR reactor with advanced robotic equipment; Cualificacion segun metodologia PDI de EPRI de te cnicas UT Phased Array para la inspeccion de vasijas de reactor PWR con eq uipos roboticos avanzados

    Energy Technology Data Exchange (ETDEWEB)

    Gadea, J. R.; Gonzalez, P.; Fernandez, F.

    2014-07-01

    The techniques and procedures qualified in the program EPRI PDI are directly applicable in plants whose reference code is ASME XI - specifically the Appendix VIII-, mainly USA and countries in which it is established American PWR technology. While countries with reactors in operation technology ABB (Sweden) or type VVER (Finland and Eastern countries) requires a qualification of specific technical type ENIQ, PDI qualification is a valuable reference since it allows to deal with such qualifications with guarantees. (Author)

  14. Experience and evaluation of advanced on-line core monitoring system 'BEACON' at IKATA site

    International Nuclear Information System (INIS)

    Shikoku Electric Power Company installed BEACON core monitoring system into IKATA unit 3 in May 1994. During its first cycle of core operation, various operational data were obtained including data of some anomalous reactor conditions introduced for the test objective of the plant start-up. This paper presents the evaluation of the BEACON system capability based on this experience. The system functions such as core monitoring and anomaly detection, prediction of future reactor conditions and increased efficiency of core management activities are discussed. Our future plan to utilize the system is also presented. (authors)

  15. The PWR programme

    International Nuclear Information System (INIS)

    For fueling the PWR type reactors two types of fuel were developed: the UO2 and mixed oxide fuels. To satisfy the demand of the operators of UO2-fuelled power plants a specific industrial organization has been established by Cogema and Framatome: Framagema supplies the technical expertise and sells the fuel; FBFC (Societe Franco-Belge de Fabrication Combustible) is manufacturing the fuel by using particularly the zirconium components produced by Zircotube and Cezus. By making possible the recycling of the materials recovered from the spent fuel reprocessing the MOX (mixed oxide fuels) technology represents an important venture for the future electronuclear sector. To implement this project Cogema created together with Belgonucleaire (the administrator of the Dessel manufacture plant) the GIE COMMOX, in charge with marketing of this fuel. On the other side Cogema which produces MOX in its facility at Cadarache, is at present building the plant at Melox of a capacity of 120 tonnes/year. After presenting the present situation with UO2 and MOX fuels the paper ends with considerations concerning the future fuels and fuels for future and further future reactors

  16. Navy lifts veil on PWR research

    International Nuclear Information System (INIS)

    The author describes the experience of Rolls Royce in developing nuclear reactors for the Navy. Reference is made to the commissioning of HMS Sceptre in February 1978, Britain's 14th nuclear submarine. This event coincided with a decision to lift the veil somewhat on a Research and Development programme that has remained secret for nearly 20 years. Factors that have inhibited progress in this field are mentioned. One of these factors has been the high cost of marine nuclear propulsion systems, tending to limit interest to very large vessels or some special purpose craft. Another factor has been slowness to develop universally acceptable safety criteria, to allow for free and ready access of nuclear vessels to ports. A third factor has been the military origins of much of the development work. A new factor that has arisen recently is the development of the Westinghouse PWR (pressurised water reactor) for marine use in the UK. This has involved collaboration with the US Westinghouse Electric Corporation. Rolls Royce and Associates were chosen to manage this work, which is here described, including the first PWR to be designed and built in Britain and incorporated into a submarine (HMS Vulcan). Much of the design work has been concerned with development of the reactor core and increasing the endurance of the vessel between refuellings. Another aspect was less noise and vibration. Costs of this work are stated, and new test facilities are described. (U.K.)

  17. Research on Operation and Control Strategy of 600MW PWR in Load Follow

    International Nuclear Information System (INIS)

    600MW Pressurized Water Reactor (PWR) is designed to operate in Constant Axial Offset Control (CAOC) strategy with base load originally. By calculations over a typical load follow scenario '12-3-6-3 (100-50-100%FP) via the CASMO-4E and SIMULATE-3 package, values of core operating parameter have been examined. With the progress of the nuclear power industry, advanced reactors are considered to have a good performance in load follow, economy and flexibility. Under the premise of fuel loading and structural dimensions unchanged, two independent control rod groups M and AO are used in 600MW pressurized water reactor to provide fine control of both the core reactivity and axial power distribution, which is named ' Improved G strategy .' The influences of different control rod distributions, composition materials, and overlap steps had in power changes have been examined in a comparative study to choose the optimal one.Then we simulate a range of load follow scenarios of the redesigned 600MW core without adjusting soluble boron concentration in the begin, middle and end of first cycle. This paper additionally demonstrated the moderator temperature coefficient and shutdown margin values of the reactor in Improved G strategy to compare with the thermal safety design criteria. It's demonstrated that adequate adjustment of control rod groups enable the core to perform load follow through Improved G strategy in 80% of cycle and save a large volume of liquid effluent particularly toward the end of cycle

  18. Core Competencies in Advanced Training: What Supervisors Say about Graduate Training

    Science.gov (United States)

    Nelson, Thorana S.; Graves, Todd

    2011-01-01

    In an attempt to identify needed mental health skills, many professional organizations have or are in the process of establishing core competency standards for their professions. The AAMFT identified 128 core competencies for the independent practice of MFT. The aim of this study was to learn the opinions of AAMFT Approved Supervisors as to how…

  19. Advances in Research on Modern Agricultural Development in Grain Production Core Area of China

    Institute of Scientific and Technical Information of China (English)

    Yan; LIU

    2015-01-01

    Grain production core area is key region of modern agricultural development in China. Through summarizing related literature about grain production area and modern agricultural development researches both at home and abroad,it obtained characteristics and existing problems in the modern agricultural development of the grain production core area. It is found that there are many research perspectives in modern agricultural development of the grain production core area. On the basis of analyzing the grain production core area and connotation,mode and evaluation of the modern agricultural development,it is concluded that further study should be carried out for adopting which development mode and how to make evaluation,so as to provide theoretical guidance for balanced development of modern agriculture in grain production core area of different regions.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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. New methods development for SCORPIO-VVER core monitoring systems to address advanced VVER 440 fuel types

    International Nuclear Information System (INIS)

    With introduction of advanced design fuel with Gd burnable absorber to Czech and Slovak VVER 440 reactors SCORPIO-VVER CMS faces new requirements and challenges. New methodology and tools had to be developed in the area of core design (neutron physics, core thermal hydraulics and fuel thermal mechanics) to properly model and address new design features of Gadolinium bearing fuel of Gd 1 and Gd 2 type. These methods have to be adapted for implementation in SCORPIO-VVER CMS. The paper provides comprehensive list of requirements and open questions, which need to be properly addressed and clearly defined prior to major innovation of the system commence. All related fields are being step by step re-evaluated (neutron physics, thermal-hydraulics and fuel thermal-mechanics). Pin power determination methodology had to be improved. Higher geometrical complexity on upper and lower core ends, particularly for transition cores with different fuel types, led to change in axial nodalization. More stringent fuel related limits (design criteria, lover margins) with higher burn-up and required unit (fuel) maneuverability require new calculation strategy in fuel conditioning/de-conditioning PCI supervising module PES. Validation of simplified simulator for HiBu domain is under development using FEMAXI code and new 2D-3D tool development for CEZ utility. Conclusions of the paper concentrate on mid term and long term innovation plans for core and fuel operation reliability crucial systems. (Author)

  3. A practical method for optimization of fuel management of PWR

    International Nuclear Information System (INIS)

    A practical method for simulation of fuel management optimization of PWR cores with two-dimensional model is described. The general objective of the optimization is to choose a set of refuelling arrangement schemes, which will produce the maximum economic profit on condition that it meets the safety criteria of PWR. It oftern requires quite a lot of computer time to simulate the optimized schemes. An effective and acceptable optimization strategy, two-step search method, has been developed. The first step of algorithm consists of several approaches based on the information avilable and the past experiences with refuelling. The second step allows a further improvement of the previously determined optimum schemes. The maximum radial power peaking factor, Wp, is defined as the objective function. Several physical criteria are examined to propose the constraints. The main intention is to minimize, the objective function Wp, subjected to various constraints. Hence a computer program, 2DFEOF in FORTRAN 77, was developed. Some calculations were done for a typical PWR core on an IBM-4341 computer. The satisfactory results were obtained at reasonable low computational costs. It spent nearly 9 mins CPU time for 3 fuel cycles with a 1/8 core configuration

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

    Science.gov (United States)

    Thiollay, Nicolas; Di Salvo, Jacques; Sandrin, Charlotte; Soldevila, Michel; Bourganel, Stéphane; Fausser, Clément; Destouches, Christophe; Blaise, Patrick; Domergue, Christophe; Philibert, Hervé; Bonora, Jonathan; Gruel, Adrien; Geslot, Benoit; Lamirand, Vincent; Pepino, Alexandra; Roche, Alain; Méplan, Olivier; Ramdhane, Mourad

    2016-02-01

    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.

  5. 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.

  6. Recent advances in the synthesis of Fe3O4@AU core/shell nanoparticles

    Science.gov (United States)

    Salihov, Sergei V.; Ivanenkov, Yan A.; Krechetov, Sergei P.; Veselov, Mark S.; Sviridenkova, Natalia V.; Savchenko, Alexander G.; Klyachko, Natalya L.; Golovin, Yury I.; Chufarova, Nina V.; Beloglazkina, Elena K.; Majouga, Alexander G.

    2015-11-01

    Fe3O4@Au core/shell nanoparticles have unique magnetic and optical properties. These nanoparticles are used for biomedical applications, such as magnetic resonance imaging, photothermal therapy, controlled drug delivery, protein separation, biosensors, DNA detection, and immunosensors. In this review, recent methods for the synthesis of core/shell nanoparticles are discussed. We divided all of the synthetic methods in two groups: methods of synthesis of bi-layer structures and methods of synthesis of multilayer composite structures. The latter methods have a layer of "glue" material between the core and the shell.

  7. Subcooled decompression analysis in PWR LOCA

    International Nuclear Information System (INIS)

    The thermo-hydraulic behavior of the coolant in the primary system of a nuclear reactor is important in the core heat transfer analysis during a hypothetical loss-of-coolant accident (LOCA). The heat transfer correlations are strongly dependent on local thermo-hydraulic conditions of the coolant. The present work allows to calculate such thermo-hydraulic behavior of the coolant during subcooled decompression in PWR LOCA by solving the mass, momentum, and energy conservation equations by the method of characteristics. Detailed studies were made on the transient coolant outflow at the pipe rupture and the effect of frictional loss and heat addition to the coolant on the decompression. Based on the studies, a digital computer code, DEPCO-MULTI, has been prepared and numerical results are compared with the ROSA (JAERI) and the LOFT (NRTS) semiscale test data with various coolant pressures, temperatures, pipe break sizes, and complexity of flow geometry. Good agreement is generally obtained

  8. Sizewell 'B' PWR reference design

    International Nuclear Information System (INIS)

    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. Penn State advanced light water reactor concept

    International Nuclear Information System (INIS)

    The accident at Three Mile Island heightened concerns over the safety of nuclear power. In response to these concerns, a research group at the Pennsylvania State University (Penn State) undertook the conceptual design of an advanced light water reactor (ALWR) under sponsorship of the US Dept. of Energy (DOE). The design builds on the literally hundreds of years worth of experience with light water reactor technology. The concept is a reconfigured pressurized water reactor (PWR) with the capability of being shut down to a safe condition simply by removing all ac power, both off-site and on-site. Using additional passively activated heat sinks and replacing the pressurizer with a pressurizing pump system, the concept essentially eliminates the concerns of core damage associated with a total station blackout. Evaluation of the Penn State ALWR concept has been conducted using the EPRI Modular Modeling System (MMS). Results show that a superior response to normal operating transients can be achieved in comparison to the response with a conventional PWR pressurizer. The DOE-sponsored Penn State ALWR concept has evolved into a significant reconfiguration of a PWR leading to enhanced safety characteristics. The reconfiguration has touched a number of areas in overall plant design including a shutdown turbine in the secondary system, additional passively activated heat sinks, a unique primary side pressurizing concept, a low pressure cleanup system, reactor building layout, and a low power density core design

  10. Borssele PWR noise: measurements, analysis and interpretation

    International Nuclear Information System (INIS)

    In the Borssele reactor - a 450 MWe PWR - reactor noise measurements have been performed during four fuel cycles. Measurements were made with a set of ex-core neutron detectors, on one occasion an in-core displacement transducer, and with primary coolant pressure sensors. Digital analysis was applied, where the most powerful tool was the computer programme FAST, which computes auto and cross power spectra for all combinations from a set of many simultaneously recorded signals. Analyses of neutronic signals show a reactivity noise peak at 9.2 Hz, core barrel motion peaks at about 12 and 15 Hz, a damped oscillation at about 2 Hz. Results are given for begin and end of each fuel cycle. The r.m.s. value of the low frequency noise appears to depend linearly on the boron concentration over a wide range. Also some results of primary coolant pressure noise are presented, with coherent peaks below 15 Hz and incoherent peaks above. The second part of the paper describes an alternative way of analyzing and interpreting noise spectra, namely attempts to decompose the neutronic power spectra into physical components, using the information present in the CPSD's of all detector combinations. The components are characterised by their detector position dependency. Effects considered are: uncorrelated noise, global reactivity noise, core motion attenuation noise, and a possible coupling between reactivity and core motion. Results show excellent separation into reactivity and core motion components with possibilities to separate overlapping peaks. Weak peaks become more easily detectable. At low frequencies the decomposition of the spectra is not yet complete, however. (author)

  11. Core design and fuel management

    International Nuclear Information System (INIS)

    This lecture reports on the experience with operating power reactors rather than on theoretical methods of core design. The principles of fuel management will be demonstrated in more detail for a PWR while the basis of core analysis will be shown in a greater extent for a BWR. (orig.)

  12. Advanced sodium cooled reactor cores having thorium blankets for effective burning of transuranic nuclides

    International Nuclear Information System (INIS)

    In this paper, a design concept of 400 MWe sodium cooled fast reactor (SFR) cores having thorium blankets for effective burning of TRU (Transuranics) from LWR spent fuel is described. Specifically, we considered two recycling options of thorium blankets : 1) no recycling and 2) fully recycling. The thorium blankets are loaded in the axially central regions of the core regions and their axial heights are adjusted so as to increase TRU burning rate and to reduce burnup reactivity swing. Also, we analyzed the performances of the cores having different fuel management batch sizes and different recycling options for the searched core configuration. The results show that the axial thorium blankets with no recycling option can be effectively used to increase TRU burning rate with a significant reduction of burnup reactivity swing in comparison with typical SFR burner cores having no blankets while the recycling of thorium blanket degrades TRU burning rate and burnup reactivity swing but it leads to a reduction of sodium void worth and more negative Doppler coefficient. (author)

  13. Recent advances in the synthesis of Fe3O4@AU core/shell nanoparticles

    International Nuclear Information System (INIS)

    Fe3O4@Au core/shell nanoparticles have unique magnetic and optical properties. These nanoparticles are used for biomedical applications, such as magnetic resonance imaging, photothermal therapy, controlled drug delivery, protein separation, biosensors, DNA detection, and immunosensors. In this review, recent methods for the synthesis of core/shell nanoparticles are discussed. We divided all of the synthetic methods in two groups: methods of synthesis of bi-layer structures and methods of synthesis of multilayer composite structures. The latter methods have a layer of “glue” material between the core and the shell. - Highlights: • Fe3O4 nanoparticles are promising for biomedical applications but have some disadvantages. • Covering Fe3O4 nanoparticles with Au shell leads to better stability and biocompatibility. • Core/shell nanoparticles are widely used for biomedical applications. • There are two types of Fe3O4@Au core/shell nanoparticles structures: bi-layer and multilayer composite. • Different synthetic methods enable production of nanoparticles of different sizes

  14. Condensate purification in PWR reactors

    International Nuclear Information System (INIS)

    The recommendations made by the VGB task group on 'condensate purification for PWR reactors' 1976 are discussed in detail. Techniques and circuiting possibilities of condensate purification for BBR steam generators (forced circulation) and KWU steam generators (U tube with blow-down) are mentioned. (HP)

  15. PWR AXIAL BURNUP PROFILE ANALYSIS

    International Nuclear Information System (INIS)

    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)

  16. PWR AXIAL BURNUP PROFILE ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    J.M. Acaglione

    2003-09-17

    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&W 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).

  17. New methods development for SCORPIO-VVER core monitoring systems to address advanced VVER 440 fuel types

    International Nuclear Information System (INIS)

    With the introduction of advanced design fuel with the Gd burnable absorber to Czech and Slovak VVER 440 reactors, the SCORPIO-VVER core monitoring system (CMS) faces new requirements and challenges. New methodology and tools had to be developed in the area of core design (neutron physics, core thermal hydraulics and fuel thermal mechanics) to properly model and address new design features of Gadolinium bearing fuel of Gd 1 and Gd 2 type. These methods have to be adapted for implementation in SCORPIO-VVER CMS. The paper provides a comprehensive list of requirements and open questions that need to be properly addressed and clearly defined prior to starting any major system innovation. Validation of a simplified simulator for the HiBu domain is under development using FEMAXI code and new 2D-3D tool development for the CEZ utility. Conclusions of the paper concentrate on mid-term and long-term innovation plans for core and fuel operation reliability systems

  18. A pressure drop model for PWR grids

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Dong Seok; In, Wang Ki; Bang, Je Geon; Jung, Youn Ho; Chun, Tae Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A pressure drop model for the PWR grids with and without mixing device is proposed at single phase based on the fluid mechanistic approach. Total pressure loss is expressed in additive way for form and frictional losses. The general friction factor correlations and form drag coefficients available in the open literatures are used to the model. As the results, the model shows better predictions than the existing ones for the non-mixing grids, and reasonable agreements with the available experimental data for mixing grids. Therefore it is concluded that the proposed model for pressure drop can provide sufficiently good approximation for grid optimization and design calculation in advanced grid development. 7 refs., 3 figs., 3 tabs. (Author)

  19. Basic study on PWR plant behavior under the condition of severe accident

    International Nuclear Information System (INIS)

    In this paper, we report on the core cooling effect by natural circulation cooling of the primary cooling system in all core cooling function loss accidents caused by SBO in PWR plant compared with BWR. We also report on the core cooling effect by using air as the final heat sink in place of the seawater by opening the main steam valve of the steam generator. On the other hand, we discuss the behavior of PWR plant in the most serious case that the damage such as LOCA is caused by earthquake and that SBO due to the subsequent tsunami causes the reactor isolation and all function of reactor core cooling system loss. That is the case that LOCA and SBO occur in superimposed manner. We can show the results from the simulation experiments that, in PWR plant, even if it is fell into the reactor core cooling function loss due to SBO, natural circulation cooling can keep the reactor core cool down as long as the feed water is supplied to SG by the turbine-driven auxiliary feed-water pump and also that the cooling effect of even more is expected by ensuring the heat-pass to the atmosphere by opening the main steam valve. We also clarify the plant behaviors under the condition that LOCA and SBO occur in superimposed manner in PWR through the simulation experiments. (author)

  20. Design study of long-life PWR using thorium cycle

    Science.gov (United States)

    Subkhi, Moh. Nurul; Su'ud, Zaki; Waris, Abdul

    2012-06-01

    Design study of long-life Pressurized Water Reactor (PWR) using thorium cycle has been performed. Thorium cycle in general has higher conversion ratio in the thermal spectrum domain than uranium cycle. Cell calculation, Burn-up and multigroup diffusion calculation was performed by PIJ-CITATION-SRAC code using libraries based on JENDL 3.2. The neutronic analysis result of infinite cell calculation shows that 231Pa better than 237Np as burnable poisons in thorium fuel system. Thorium oxide system with 8% 233U enrichment and 7.6˜ 8% 231Pa is the most suitable fuel for small-long life PWR core because it gives reactivity swing less than 1% Δk/k and longer burn up period (more than 20 year). By using this result, small long-life PWR core can be designed for long time operation with reduced excess reactivity as low as 0.53% Δk/k and reduced power peaking during its operation.

  1. Advanced Core Design And Fuel Management For Pebble-Bed Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

    2004-10-01

    A method for designing and optimizing recirculating pebble-bed reactor cores is presented. At the heart of the method is a new reactor physics computer code, PEBBED, which accurately and efficiently computes the neutronic and material properties of the asymptotic (equilibrium) fuel cycle. This core state is shown to be unique for a given core geometry, power level, discharge burnup, and fuel circulation policy. Fuel circulation in the pebble-bed can be described in terms of a few well?defined parameters and expressed as a recirculation matrix. The implementation of a few heat?transfer relations suitable for high-temperature gas-cooled reactors allows for the rapid estimation of thermal properties critical for safe operation. Thus, modeling and design optimization of a given pebble-bed core can be performed quickly and efficiently via the manipulation of a limited number key parameters. Automation of the optimization process is achieved by manipulation of these parameters using a genetic algorithm. The end result is an economical, passively safe, proliferation-resistant nuclear power plant.

  2. Recent advances on thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alexandro S., E-mail: alexandrossilva@ifba.edu.br [Instituto Federal de Educacao, Ciencia e Tecnologia da Bahia (IFBA), Vitoria da Conquista, BA (Brazil); Mazaira, Leorlen Y.R., E-mail: leored1984@gmail.com, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (INSTEC), La Habana (Cuba); Dominguez, Dany S.; Hernandez, Carlos R.G., E-mail: alexandrossilva@gmail.com, E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil). Programa de Pos-Graduacao em Modelagem Computacional; Lira, Carlos A.B.O., E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil)

    2015-07-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal-hydraulic characteristics. In this article, it was performed the thermal-hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a FCC (Face Centered Cubic) cell with the half height of the core, with 21 layers and 95 pebbles. The input data used were taken from the thermal-hydraulic IAEA Bechmark. The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  3. REWET, PWR LOCA accident experiments

    International Nuclear Information System (INIS)

    1 - Description of test facility: The REWET-II facility was designed for the investigation of the reflooding phase of a LOCA. The main design principle is the accurate simulation of the rod bundle geometry and the primary system elevations. This is necessary in order to have the correct flow channels and hydrostatic pressures for the reflooding process. The reactor vessel is simulated by a stainless steel U-tube construction consisting of downcomer, lower plenum, core and upper plenum. The primary loops contain a pipe simulating the broken loop and a connection line between the upper plenum and the downcomer simulating five intact loops. The containment is simulated by a pressure vessel (not in scale). Steam generators and primary pumps are simulated with flow resistances. The ECC-water can be injected to the downcomer and/or to the upper plenum by a pump or from an accumulator. All the elevation in the reactor vessel simulator are scaled to 1:1 (except the reactor upper head). The scale of the volumes and flow areas is 1:2333 referring to the number of the fuel rod simulators in the facility and the fuel rods in the reference reactor. The rod bundle is either in a hexagonal shroud, the inside distance of the opposite walls is 54.3 mm and the wall thickness 2 mm, or in a round shroud, the inner diameter is 66.0 mm and the wall thickness 2 mm. The simulation of the fuel-rod bundle consists of 19 indirectly electrically heated simulator rods. The heating coils are inside stainless steel claddings in magnesium oxide insulation. The heated length, the outer diameter and the lattice pitch of the fuel-rod simulators as well as the number (= 10) and construction of the rod bundle spacers are the same as in the reference reactor. The upper ends of the rods are attached to the upper tie plate. 2 - Description of test: Pressurized water reactors in use in Finland reactors have certain unique features which make them different from most other PWR designs. The 6 horizontal

  4. Design Development and Verification of a System Integrated Modular PWR

    International Nuclear Information System (INIS)

    An advanced PWR with a rated thermal power of 330 MW has been developed at the Korea Atomic Energy Research Institute (KAERI) for a dual purpose: seawater desalination and electricity generation. The conceptual design of SMART ( System-Integrated Modular Advanced ReacTor) with a desalination system was already completed in March of 1999. The basic design for the integrated nuclear desalination system is currently underway and will be finished by March of 2002. The SMART co-generation plant with the MED seawater desalination process is designed to supply forty thousand (40,000) tons of fresh water per day and ninety (90) MW of electricity to an area with approximately a ten thousand (100,000) population or an industrialized complex. This paper describes advanced design features adopted in the SMART design and also introduces the design and engineering verification program. In the beginning stage of the SMART development, top-level requirements for safety and economics were imposed for the SMART design features. To meet the requirements, highly advanced design features enhancing the safety, reliability, performance, and operability are introduced in the SMART design. The SMART consists of proven KOFA (Korea Optimized Fuel Assembly), helical once-through steam generators, a self-controlled pressurizer, control element drive mechanisms, and main coolant pumps in a single pressure vessel. In order to enhance safety characteristics, innovative design features adopted in the SMART system are low core power density, large negative Moderator Temperature Coefficient (MTC), high natural circulation capability and integral arrangement to eliminate large break loss of coolant accident, etc. The progression of emergency situations into accidents is prevented with a number of advanced engineered safety features such as passive residual heat removal system, passive emergency core cooling system, safeguard vessel, and passive containment over-pressure protection. The preliminary

  5. Computer analyses on loop seal clearing experiment at PWR PACTEL

    International Nuclear Information System (INIS)

    Highlights: • Code analyses of loop seal clearing experiment with PWR PACTEL are introduced. • TRACE and APROS system codes are used in the analyses. • Main events of the experiment are well predicted with both codes. • Discrepancies are observed on the secondary side and in the core region. • Loop seal clearing phenomenon is well simulated with both codes. - Abstract: Water seal formation in the loop seal in pressurized water reactors can occur during a small or intermediate break loss-of-coolant accident, causing temporary fuel overheating. Quantification of the accuracy of overheating prediction is of interest in the best-estimate safety analyses, even though the peak cladding temperatures due to the water seal formation in the loop seal seldom approach acceptance criteria as such. The aim of this study was to test and evaluate the accuracy with which the thermal–hydraulic system code nodalizations of the PWR PACTEL predict loop seal clearing in a small break loss-of-coolant-accident test performed with the PWR PACTEL facility. PWR PACTEL is a thermal–hydraulic test facility with two loops and vertical inverted U-tube steam generators. Post-test simulations were performed with the TRACE and APROS system codes. In the post-test simulations, the main events of the transient such as the decrease in the core water level, depressurization of the primary circuit, and the behavior of the water seal formation and clearing in the loop seal were predicted satisfactorily by both codes. However, discrepancies with the experiment results were observed in the analyses with both codes, for example the core temperature excursions were halted too early and the peak temperature predictions were too low. The core water level increase caused by loop seal clearing was overestimated with both codes, and the pressure and temperature were overestimated on the secondary side of the steam generators. Loop Seal 2 was evidently cleared out while Loop Seal 1 remained closed

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

    International Nuclear Information System (INIS)

    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% U3 O8 and 6% separative work units if recycle is assumed only after the fifth operation cycle of the thermal reactors. (author)

  7. Serious accidents of PWR type reactors for power generation

    International Nuclear Information System (INIS)

    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.)

  8. PWR plant transient analyses using TRAC-PF1

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Fiel, Joao Claudio B., E-mail: fiel@ime.eb.br [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Departamento de Engenharia Nuclear

    2015-07-01

    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 {sup 235} U {sub 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{sub inf}, generated by MCNPX code calculations, show a difference of less than 0.7 percent. (author)

  10. Neutronic performance of uranium nitride composite fuels in a PWR

    International Nuclear Information System (INIS)

    Highlights: • Survey and sensitivity assembly level studies for uranium nitride composite fuels. • Composites harden the neutron spectrum and decrease the worth of control rods. • Moderator temperature coefficient is more negative, soluble boron coefficient is less negative. • Similar equilibrium core power peaking and reactivity coefficient when compared to UO2. • Illustrates “do no harm” in evaluation of candidate accident tolerant fuels. - Abstract: Uranium mononitride (UN) based composite nuclear fuels may have potential benefits in light water reactor applications, including enhanced thermal conductivity and increased fuel density. However, uranium nitride reacts chemically when in contact with water, especially at high temperatures. To overcome this challenge, several advanced composite fuels have been proposed with uranium nitride as a primary phase. The primary nitride phase is “shielded” from water by a secondary phase, which would allow the potential benefits of nitride fuels to be realized. This work is an operational assessment of four different candidate composite materials. We considered uranium dioxide (UO2) and UN base cases and compared them with the candidate composite UN-based fuels. The comparison was performed for nominal conditions in a reference PWR with Zr-based cladding. We assessed the impact of UN porosity on the operational performance, because this is a key sensitivity parameter. As composite fuels, we studied UN/U3Si5, UN/U3Si2, UN/UB4, and UN/ZrO2. In the case of UB4, the boron content is 100% enriched in 11B. The proposed zirconium dioxide (ZrO2) phase is cubic and yttria-stabilized. In all cases UN is the primary phase, with small fractions of U3Si5, U3Si5, UB4, or ZrO2 as a secondary phase. In this analysis we showed that two baseline nitride cases at different fractions of theoretical density (0.8 and 0.95) generally bound the neutronic performance of the candidate composite fuels. Performance was comparable with

  11. Advanced core physics and thermal hydraulics analysis of boiling water reactors using innovative fuel concepts

    OpenAIRE

    Winter, Dominik

    2014-01-01

    The economical operation of a boiling water reactor (BWR) is mainly achieved by the axially uniform utilization of the nuclear fuel in the assemblies which is challenging because the neutron spectrum in the active reactor core varies with the axial position. More precisely, the neutron spectrum becomes harder the higher the position is resulting in a decrease of the fuel utilization because the microscopic fission cross section is smaller by several orders of magnitude. In this work, the use ...

  12. Influence of high dose irradiation on core structural and fuel materials in advanced reactors

    International Nuclear Information System (INIS)

    The IAEA International Working Group on Fast Reactors (IWGFR) periodically organizes meeting to discuss and review important aspects of fast reactor technology. The fifth meeting held in Obninsk, Russian Federation, 16-19 June 1997, was devoted to the influence of high dose irradiation on the mechanical properties of reactor core structural and fuel materials. The proceedings includes the papers submitted at this meeting each with a separate abstract

  13. Core design and fuel cycle of advanced fast reactor with sodium coolant

    International Nuclear Information System (INIS)

    A perspective sodium reactor is under development in Russia nowadays. Initially, power level of 1800 MW (el.) was considered for this reactor. However, owing to many reasons, in particular, for transportability of the main plant by railway, the reactor power was later reduced to 1200 MW (el.). At the same time the base of the concept for the choice of the core parameters remained the same as for the 1800 MW power, including the following: - low core specific power resulting in a decrease of the fuel lifetime and, consequently, a smaller annual consumption of fuel elements; - enhancement of inherent self-protection: ensuring the sodium void reactivity effect (SVR) close to zero and a minimum reactivity margin for burnup; - ensuring the reactor operation in different patterns of the closed fuel cycle organization: the use of plutonium from thermal reactor with and without MA for the first loading, recycling the own plutonium with/without breeding, burnup of own MA, etc. Basic characteristics of the core of BN-1200 reactor approved for the current phase of designing have been reported. (author)

  14. Core Principles and Test Item Development for Advanced High School and Introductory University Level Food Science

    Science.gov (United States)

    Laing-Kean, Claudine A. M.

    2010-01-01

    Programs supported by the Carl D. Perkins Act of 2006 are required to operate under the state or national content standards, and are expected to carry out evaluation procedures that address accountability. The Indiana high school course, "Advanced Life Science: Foods" ("ALS: Foods") operates under the auspices of the Perkins Act. However, no broad…

  15. Enriched Gadolinium as burnable absorber for PWR

    International Nuclear Information System (INIS)

    This paper is a summary of a master of thesis work in reactor physics made by Ola Seveborn. The work was done at Vattenfall Braensle AB and Ola was guided through the work by the corresponding author of this paper. The results presented are calculations for Ringhals 3, which is a Westinghouse 3-loop PWR within the Vattenfall Group. The fuel is characterized by 17x17 assemblies of AFA type containing 3.80-3.95 w/o 235U and 8 rods containing 2 w/o Gadolinium with an enrichment of 70 w/o 157Gd. The calculations were performed with the Studsvik-Scandpower code package based on the CASMO-4 lattice code and the SIMULATE-3 nodal code. The results are compared to the corresponding calculations for fuel with 5 w/o gadolinium with natural isotopic constitution. The depletion of the cores was done separately for the reference and enriched case. The results show that the gains in average for the five cycles studied are about 70 EFPH per cycle. This is an effect of the lower gadolinium content needed. Also less parasitic absorption of enriched gadolinium in the end of the fuel life contributes to the increased cycle lengths. The abruptly increased reactivity and internal power peaking factor around 10 MWd/kgU do not affect the core design negatively. (authors)

  16. Overview of PWR chemistry options

    Energy Technology Data Exchange (ETDEWEB)

    Nordmann, F.; Stutzmann, A.; Bretelle, J.L. [Electricite de France, Central Labs. (France)

    2002-07-01

    EDF Central Laboratories, in charge of engineering in chemistry, of defining the chemistry specifications and studying the operation feedback and improvement for 58 PWR units, have the opportunity to evaluate many options of operation developed and applied all around the world. Thanks to these international relationships and to the benefit of a large feedback from many units, some general evaluation of the various options is discussed in this paper. (authors)

  17. Advances in high-pressure mineral physics: from the deep mantle to the core

    OpenAIRE

    Ohtani, Eiji; Andrault, Denis; Stixrude, Lars; Wang, Yanbin; Asimow, Paul D.

    2009-01-01

    Mineral physics studies provide basic information on physical, chemical, thermodynamic and transport properties of constituents of the Earth’s interior. This information, combined with other geophysical observations, helps constrain the structure and dynamics of the Earth. This Physics of the Earth and Planetary Interiors special issue is a collection of the experimental, computational and theoretical research and review papers relating to recent “Advances in Mineral Phys...

  18. PWR secondary water chemistry guidelines: Revision 3

    International Nuclear Information System (INIS)

    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)

  19. Advances in the development of liquid-core waveguides for IR applications

    Science.gov (United States)

    Meister, Joerg; Jung, Roland; Diemer, Stefan; Haisch, Michael; Fuss, Werner; Hering, Peter

    1996-04-01

    With the development of infrared transmitting fibers, medical applications such as minimally invasive surgery are becoming feasible. In particular we investigate liquid core waveguides with an Er:YAG laser at 2.94 micrometer. Because of their advantages like variability in diameter, high flexibility, and mechanical stability, liquid core waveguides appear to be an alternative to conventional IR waveguides. In this work we present two types of liquid CCl4 filled lightguides that have been developed with plastic tube and quartz capillary as cladding. The former with an inner diameter of 1.6 mm showed an attenuation of 2.6 dB/m at 2.94 micrometer. For the quartz glass capillary with an inner diameter of 550 micrometers an attenuation of approximately 4.8 dB/m was determined in first experimental results. Due to the great flexibility and the high mechanical stability of both lightguides, bending radii below 10 mm are possible. Transmission losses depending on bending radii are discussed. A comparison between measurements with an IR-spectrometer and an Er:YAG laser shows that a minimum transmission loss of 2 dB/m can be achieved.

  20. Uncertainty Evaluation of Reactivity Coefficients for a large advanced SFR Core Design

    International Nuclear Information System (INIS)

    Sodium Cooled Fast Reactors are currently being reshaped in order to meet Generation IV goals on economics, safety and reliability, sustainability and proliferation resistance. Recent studies have led to large SFR cores for a 3600 MWth power plants, cores which exhibit interesting features. The designs have had to balance between competing aspects such as sustainability and safety characteristics. Sustainability in neutronic terms is translated into positive breeding gain and safety into rather low Na void reactivity effects. The studies have been done on two SFR concepts using oxide and carbide fuels. The use of the sensitivity theory in the ERANOS determinist code system has been used. Calculations have been performed with different sodium evaluations: JEF2.2, ERALIB-1 and the most recent JEFF3.1 and ENDF/B-VII in order to make a broad comparison. Values for the Na void reactivity effect exhibit differences as large as 14% when using the different sodium libraries. Uncertainties due to nuclear data on the reactivity coefficients were performed with BOLNA variances-covariances data, the Na Void Effect uncertainties are near to 12% at 1 σ. Since, the uncertainties are far beyond the target accuracy for a design achieving high performance, two directions are envisaged: the first one is to perform new differential measurements or in a second attempt use integral experiments to improve effectively the nuclear data set and its uncertainties such as performed in the past with ERALIB1. (authors)

  1. Advanced core physics and thermal hydraulics analysis of boiling water reactors using innovative fuel concepts

    International Nuclear Information System (INIS)

    The economical operation of a boiling water reactor (BWR) is mainly achieved by the axially uniform utilization of the nuclear fuel in the assemblies which is challenging because the neutron spectrum in the active reactor core varies with the axial position. More precisely, the neutron spectrum becomes harder the higher the position is resulting in a decrease of the fuel utilization because the microscopic fission cross section is smaller by several orders of magnitude. In this work, the use of two fuel concepts based on a mixed oxide (MOX) fuel and an innovative thorium-plutonium (ThPu) fuel is investigated by a developed simulation model encompassing thermal hydraulics, neutronics, and fuel burnup. The main feature of these fuel concepts is the axially varying enrichment in plutonium which is, in this work, recycled from spent nuclear fuel and shows a high fission fraction of the absorption cross section for fast incident neutron energies. The potential of balancing the overall fuel utilization by an increase of the fission rate in the upper part of the active height with a combination of the harder spectrum and the higher fission fraction of the absorption cross section in the BWR core is studied. The three particular calculational models for thermal hydraulics, neutronics, and fuel burnup provide results at fuel assembly and/or at core level. In the former case, the main focus lies on the thermal hydraulics analysis, fuel burnup, and activity evolution after unloading from the core and, in the latter case, special attention is paid to reactivity safety coefficients (feedback effects) and the optimization of the operational behavior. At both levels (assembly and core), the isotopic buildup and depletion rates as a function of the active height are analyzed. In addition, a comparison between the use of conventional fuel types with homogeneous enrichments and the use of the innovative fuel types is made. In the framework of the simulations, the ThPu and the MOX

  2. 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)

    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.)

  3. PWR type reactor

    International Nuclear Information System (INIS)

    Coolant discharging windows disposed to a control rod cluster guide tube are distributed in a region between the height of the lower end of a coolant exit nozzle and the height of the lower nozzle of an upper reactor core support column. The flow of coolants in the lateral direction toward an exit nozzle does not flow backwardly from the discharging windows to the inside of the control rod cluster guide tube, and the flow of coolants in the control rod cluster guide tube is discharged from each of the coolant discharging windows to the outside directly and rapidly while forming branched streams. As a result, the flow rate of coolants passing through a continuous portion is greatly reduced, and the flow rate of coolants in the direction traversing the control rods is greatly reduced. Accordingly, fluid vibrations for all the control rod clusters is reduced to reduce abrasion and the thickness reduction of the walls of a guide plate of the control rod cluster guide tube caused by contact with the control rods. (N.H.)

  4. Advancing Nucleosynthesis in Self-consistent, Multidimensional Models of Core-Collapse Supernovae

    CERN Document Server

    Harris, J Austin; Chertkow, Merek A; Bruenn, Stephen W; Lentz, Eric J; Messer, O E Bronson; Mezzacappa, Anthony; Blondin, John M; Marronetti, Pedro; Yakunin, Konstantin N

    2014-01-01

    We investigate core-collapse supernova (CCSN) nucleosynthesis in polar axisymmetric simulations using the multidimensional radiation hydrodynamics code CHIMERA. Computational costs have traditionally constrained the evolution of the nuclear composition in CCSN models to, at best, a 14-species $\\alpha$-network. Such a simplified network limits the ability to accurately evolve detailed composition, neutronization and the nuclear energy generation rate. Lagrangian tracer particles are commonly used to extend the nuclear network evolution by incorporating more realistic networks in post-processing nucleosynthesis calculations. Limitations such as poor spatial resolution of the tracer particles, estimation of the expansion timescales, and determination of the "mass-cut" at the end of the simulation impose uncertainties inherent to this approach. We present a detailed analysis of the impact of these uncertainties on post-processing nucleosynthesis calculations and implications for future models.

  5. Recent advances in the theoretical modeling of pulsating low-mass He-core white dwarfs

    CERN Document Server

    Córsico, A H; Calcaferro, L M; Serenelli, A M; Kepler, S O; Jeffery, C S

    2016-01-01

    Many extremely low-mass (ELM) white-dwarf (WD) stars are currently being found in the field of the Milky Way. Some of these stars exhibit long-period nonradial $g$-mode pulsations, and constitute the class of ELMV pulsating WDs. In addition, several low-mass pre-WDs, which could be precursors of ELM WDs, have been observed to show short-period photometric variations likely due to nonradial $p$ modes and radial modes. They could constitute a new class of pulsating low-mass pre-WD stars, the pre-ELMV stars. Here, we present the recent results of a thorough theoretical study of the nonadiabatic pulsation properties of low-mass He-core WDs and pre-WDs on the basis of fully evolutionary models representative of these stars.

  6. QFLOOD-GT: a program for predicting PWR reflood

    International Nuclear Information System (INIS)

    A description is given of the present version of the QFLOOD-GT program for predicting the reflood stage of a large-break PWR loss-of-coolant accident. QFLOOD-GT has been developed from an earlier forced-reflood program which, using a conduction-controlled model for rewetting speed, gave good agreement with the FLECHT SEASET experiments. This earlier program has been incorporated into QFLOOD-GT as a subroutine called QFLOOD; in addition a downcomer model has been included in order to allow calculation of gravity reflood, and a computational scheme has been devised to simulate the chimney effect (the unequal distribution of inlet flow between hot and cool regions of the core). No quantitative comparisons between QFLOOD-GT predictions and integral-test data have yet been carried out, so the modelling decisions implemented in the program are at this stage unvalidated. Preliminary testing of the program has produced results which are for the most part qualitatively satisfactory. Calculations for indicative PWR conditions suggest that the chimney effect has a significant beneficial effect during PWR reflood, a conclusion in accordance with the findings of the Japanese 2D/3D experiments. (author)

  7. PWR experimental benchmark analysis using WIMSD and PRIDE codes

    International Nuclear Information System (INIS)

    Highlights: • PWR experimental benchmark calculations were performed using WIMSD and PRIDE codes. • Various models for lattice cell homogenization were used. • Multiplication factors, power distribution and reaction rates were studied. • The effect of cross section libraries on these parameters was analyzed. • The results were compared with experimental and reported results. - Abstract: The PWR experimental benchmark problem defined by ANS was analyzed using WIMSD and PRIDE codes. Different modeling methodologies were used to calculate the infinite and effective multiplication factors. Relative pin power distributions were calculated for infinite lattice and critical core configurations, while reaction ratios were calculated for infinite lattice only. The discrete ordinate method (DSN) and collision probability method (PERSEUS) were used in each calculation. Different WIMSD cross-section libraries based on ENDF/B-VI.8, ENDF/B-VII.0, IAEA, JEF-2.2, JEFF-3.1 and JENDL-3.2 nuclear data files were also employed in the analyses. Comparison was made with experimental data and other reported results in order to find a suitable strategy for PWR analysis

  8. IPSN expert appraisal programme on the chooz A 300 MWe PWR. Lessons learned by IPSN

    International Nuclear Information System (INIS)

    The closure of Chooz A PWR provided an opportunity to take samples of items that had aged in situ in conditions close to those encountered in PWR in operation over a period of 140.000 hours, which is far longer than the usual time-spans of simulated laboratory tests. 4 topics have been studied: 1) effect of radiation on reactor vessel internals, 2) dissimilar metal joints of reactor coolant system: pressurizer surge line, 3) cast parts of austeno-ferritic steel: hot and cold leg primary valves, and 4) ageing of cables in high temperatures and under irradiation. The examination of the lower internals on some baffle angle bracket and core shroud screws, subjected to varying amounts of irradiation, did not reveal any cracking or corrosion, and confirmed the saturation effect between 4 and 10 dpa for the hardening of 304 austenitic steel in the low temperature range. Expert appraisal of the dissimilar metal joints on the pressurizer surge line confirmed the existence of small fabrication defects due to high temperature cracking. Expert appraisal of the 3 valve body samples from the main section of the coolant system confirmed that -) thermal ageing of the valve body on the hot leg was more advanced than that of the cold leg valve, -) the material of the valve housing on the cold leg which, in theory, was not sensitive to ageing phenomena, exhibited unexpectedly low impact strength values. As for cables, measurements confirmed that their mechanical and electrical properties remained sufficient for them to carry out their functions. (A.C.)

  9. Global stability analysis of pressurized water reactor core nonlinear system

    International Nuclear Information System (INIS)

    Determining the global stability of a pressurized water reactor (PWR) core nonlinear system is the problem to be solved. In the paper, the core nonlinear system was modeled and the linearized model of the system was obtained via the small perturbation method. According to the distributing situation of the core nonlinearity measure in the power level range based on the equilibrium manifold, seven linear models corresponding to seven power levels respectively were chosen as local models of the core and the set of seven local models was used to approximately substitute the core system. The global stability of the PWR core nonlinear system was analyzed by utilizing Lyapunov stability theory. The calculated result shows that the core nonlinear system is globally and asymptotically stable. The modeling method of the core is effective in analyzing the global stability of a PWR core nonlinear system. (authors)

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

    International Nuclear Information System (INIS)

    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

  11. Application of the SHM and SPSM for calculations of some problems for WWER and PWR

    International Nuclear Information System (INIS)

    Results of SHM calculations applied to WWER and PWR reactors are presented to demonstrate here the potential of the Surface Harmonics Method (SHM) and the Surface Pseudo-Sources Method (SPSM). These methods indeed were developed as a solution to the problem of improvement of neutron field calculations in the cores of nuclear reactors with nonuniform lattices. (Authors)

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

    International Nuclear Information System (INIS)

    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)

  13. Proceeding of JSPS-CAS Core-University Program (CUP) on superconducting key technology for advanced fusion device

    International Nuclear Information System (INIS)

    The JSPS-CAS Core University Program (CUP) seminar on 'Superconducting Key Technology for Advanced Fusion Device' was held in Xi'an, China from October 18 to 21, 2010. This seminar was organized in the framework of the CUP in the field of plasma and nuclear fusion. This seminar honored by NIFS and ASIPP is aim to have a wide discussion on the new application and achievements on superconducting technology of nuclear fusion reactor. The superconducting technology on fusion reactor involves the fields on high current superconductor and magnet, quench protection, current control, cooling of the magnet, and reliability of large scale refrigerator. The technology on ITER high temperature superconductor current leads and the conductor test of JT-60SA are discussed in this seminar. Thirty-four oral talks and two summary talks were presented in this seminar. Total number of the participants was 34, including 12 Japanese participants. (author)

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

    International Nuclear Information System (INIS)

    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%U3O8 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, U3O8 and SWU economy with recycle is not significant when the proposed fast breeder program is considered. (Author)

  15. Global shielding analysis for the three-element core advanced neutron source reactor under normal operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    Slater, C.O.; Bucholz, J.A.

    1995-08-01

    Two-dimensional discrete ordinates radiation transport calculations were performed for a model of the three-element core Advanced Neutron Source reactor design under normal operating conditions. The core consists of two concentric upper elements and a lower element radially centered in the annulus between the upper elements. The initial radiation transport calculations were performed with the DORT two-dimensional discrete ordinates radiation transport code using the 39-neutron-group/44-gamma-ray-group ANSL-V cross-section library, an S{sub 6} quadrature, and a P{sub 1} Legendre polynomial expansion of the cross sections to determine the fission neutron source distribution in the core fuel elements. These calculations were limited to neutron groups only. The final radiation transport calculations, also performed with DORT using the 39-neutron-group/44-gamma-ray-group ANSL-V cross-section library, an S{sub l0} quadrature, and a P{sub 3} Legendre polynomial expansion of the cross sections, produced neutron and gamma-ray fluxes over the full extent of the geometry model. Responses (or activities) at various locations in the model were then obtained by folding the appropriate response functions with the fluxes at those locations. Some comparisons were made with VENTURE-calculated (diffusion theory) 20-group neutron fluxes that were summed into four broad groups. Tne results were in reasonably good agreement when the effects of photoneutrons were not included, thus verifying the physics model upon which the shielding model was based. Photoneutrons increased the fast-neutron flux levels deep within the D{sub 2}0 several orders of magnitude. Results are presented as tables of activity values for selected radial and axial traverses, plots of the radial and axial traverse data, and activity contours superimposed on the calculational geometry model.

  16. ZT-P: an advanced air core reversed field pinch prototype

    International Nuclear Information System (INIS)

    The ZT-P experiment, with a major radius of 0.45 m and a minor radius of 0.07 m, was designed to prototype the next generation of reversed field pinch (RFP) machines at Los Alamos. ZT-P utilizes an air-core poloidal field system, with precisely wound and positioned rigid copper coils, to drive the plasma current and provide plasma equilibrium with intrinsically low magnetic field errors. ZT-P's compact configuration is adaptable to test various first wall and limiter designs at reactor-relevant current densities in the range of 5 to 20 MA/m2. In addition, the load assembly design allows for the installation of toroidal field divertors. Design of ZT-P began in October 1983, and assembly was completed in October 1984. This report describes the magnetic, electrical, mechanical, vacuum, diagnostic, data acquisition, and control aspects of the machine design. In addition, preliminary data from initial ZT-P operation are presented. Because of ZT-P's prototypical function, many of its design aspects and experimental results are directly applicable to the design of a next generation RFP. 17 refs., 47 figs

  17. Advancing Pharmacogenomics Education in the Core PharmD Curriculum through Student Personal Genomic Testing.

    Science.gov (United States)

    Adams, Solomon M; Anderson, Kacey B; Coons, James C; Smith, Randall B; Meyer, Susan M; Parker, Lisa S; Empey, Philip E

    2016-02-25

    Objective. To develop, implement, and evaluate "Test2Learn" a program to enhance pharmacogenomics education through the use of personal genomic testing (PGT) and real genetic data. Design. One hundred twenty-two second-year doctor of pharmacy (PharmD) students in a required course were offered PGT as part of a larger program approach to teach pharmacogenomics within a robust ethical framework. The program added novel learning objectives, lecture materials, analysis tools, and exercises using individual-level and population-level genetic data. Outcomes were assessed with objective measures and pre/post survey instruments. Assessment. One hundred students (82%) underwent PGT. Knowledge significantly improved on multiple assessments. Genotyped students reported a greater increase in confidence in understanding test results by the end of the course. Similarly, undergoing PGT improved student's self-perceived ability to empathize with patients compared to those not genotyped. Most students (71%) reported feeling PGT was an important part of the course, and 60% reported they had a better understanding of pharmacogenomics specifically because of the opportunity. Conclusion. Implementation of PGT in the core pharmacy curriculum was feasible, well-received, and enhanced student learning of pharmacogenomics. PMID:26941429

  18. Improvement of advanced nodal method used in 3D core design system

    International Nuclear Information System (INIS)

    This paper deals with AREVA NP progress in the modelling of neutronic phenomena, evaluated through 3D determinist core codes and using 2-group diffusion theory. Our report highlights the advantages of taking into account the assembly environment in the process used for the building of the 2-group collapsed neutronic parameters, such as cross sections or discontinuity factors. The interest of the present method, developed in order to account for the impact of the environment on the above mentioned parameters, resides (i) in the very definition of a global correlation between collapsed neutronic data calculated in an infinite medium and those calculated in a 3D-geometry, and (ii) in the use of a re-homogenization method. Using this approach, computations match better with actual measurements on control rod worth. They also present smaller differences on pin by pin power values compared to the ones computed with another code considered as a reference since it relies on multigroup transport theory. (authors)

  19. Estimating probable flaw distributions in PWR steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Gorman, J.A.; Turner, A.P.L. [Dominion Engineering, Inc., McLean, VA (United States)

    1997-02-01

    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.

  20. 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)

  1. Concept of safety systems for next generation PWR (APWR+)

    International Nuclear Information System (INIS)

    The concept of the next generation PWR, which is expected to come after the APWR and is named the APWR+, is being studied, considering that the light water reactors are seemed to be dominant also in the 21st century. The APWR+ is designed to have the features of four-train safety systems, divergent emergency electrical sources, and passive core cooling system using steam generators at early stage of the Loss of Coolant Accident. The basic concept has been made, and more detailed investigation is scheduled in near future. (author)

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

    International Nuclear Information System (INIS)

    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

  3. PWR fuel management optimization using continuous particle swarm intelligence

    International Nuclear Information System (INIS)

    The objective of nuclear fuel management is to minimize the cost of electrical energy generation subject to operational and safety constraints. In the present work, a core reload optimization package using continuous version of particle swarm optimization, CRCPSO, which is a combinatorial and discrete one has been developed and mapped on nuclear fuel loading pattern problems. This code is applicable to all types of PWR cores to optimize loading patterns. To evaluate the system, flattening of power inside a WWER-1000 core is considered as an objective function although other variables such as Keff along power peaking factor, burn up and cycle length can be included. Optimization solutions, which improve the safety aspects of a nuclear reactor, may not lead to economical designs. The system performed well in comparison to the developed loading pattern optimizer using Hopfield along SA and GA.

  4. DOMINO: A fast 3D cartesian discrete ordinates solver for reference PWR simulations and SPN validation

    International Nuclear Information System (INIS)

    As part of its activity, EDF R and D is developing a new nuclear core simulation code named COCAGNE. This code relies on DIABOLO, a Simplified PN (SPN) method to compute the neutron flux inside the core for eigenvalue calculations. In order to assess the accuracy of SPN calculations, we have developed DOMINO, a new 3D Cartesian SN solver. The parallel implementation of DOMINO is very efficient and allows to complete an eigenvalue calculation involving around 300 x 109 degrees of freedom within a few hours on a single shared-memory supercomputing node. This computation corresponds to a 26-group S8 3D PWR core model used to assess the SPN accuracy. At the pin level, the maximal error for the SP5 DIABOLO fission production rate is lower than 0.2% compared to the S8 DOMINO reference for this 3D PWR core model. (authors)

  5. DOMINO: A fast 3D cartesian discrete ordinates solver for reference PWR simulations and SPN validation

    Energy Technology Data Exchange (ETDEWEB)

    Courau, T.; Moustafa, S.; Plagne, L.; Poncot, A. [EDF R and D, 1, Av du General de Gaulle, F92141 Clamart cedex (France)

    2013-07-01

    As part of its activity, EDF R and D is developing a new nuclear core simulation code named COCAGNE. This code relies on DIABOLO, a Simplified PN (SPN) method to compute the neutron flux inside the core for eigenvalue calculations. In order to assess the accuracy of SPN calculations, we have developed DOMINO, a new 3D Cartesian SN solver. The parallel implementation of DOMINO is very efficient and allows to complete an eigenvalue calculation involving around 300 x 10{sup 9} degrees of freedom within a few hours on a single shared-memory supercomputing node. This computation corresponds to a 26-group S{sub 8} 3D PWR core model used to assess the SPN accuracy. At the pin level, the maximal error for the SP{sub 5} DIABOLO fission production rate is lower than 0.2% compared to the S{sub 8} DOMINO reference for this 3D PWR core model. (authors)

  6. Thermodynamic modelling of PWR coolant

    International Nuclear Information System (INIS)

    Spinel solubilities on PWR primary circuit surfaces vary with temperature, pH and coolant H2 concentration. The available solubility data are discussed for Fe, Ni, Co and Zn oxides, and species are identified where data are very limited or absent. An equilibrium thermodynamic model is described to predict the solubility, and results are described predicting relative Fe and Ni solubility under normal operating conditions and during shutdown/startup. The relative stabilities of stoichiometric and non-stoichiometric zinc ferrite spinels are also considered. (R.P.)

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

    International Nuclear Information System (INIS)

    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

  8. 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.

  9. Development Of Advanced Sandwich Core Topologies Using Fused Deposition Modeling And Electroforming Processes

    Science.gov (United States)

    Storck, Steven M.

    New weight efficient materials are needed to enhance the performance of vehicle systems allowing increased speed, maneuverability and fuel economy. This work leveraged a multi-length-scale composite approach combined with hybrid material methodology to create new state-of-the-art additive manufactured sandwich core material. The goal of the research was to generate a new material to expands material space for strength versus density. Fused-Deposition-Modeling (FDM) was used to remove geometric manufacturing constraints, and electrodepositing was used to generate a high specific-strength, bio-inspired hybrid material. Microtension samples (3mm x 1mm with 250mum x 250mum gage) were used to investigate the electrodeposited coatings in the transverse (TD) and growth (GD) directions. Three bath chemistries were tested: copper, traditional nickel sulfamate (TNS) nickel, and nickel deposited with a platinum anode (NDPA). NDPA shows tensile strength exceeding 1600 MPa, significantly beyond the literature reported values of 60MPa. This strengthening was linked to grain size refinement into the sub-30nm range, in addition to grain texture refinement resulting in only 17% of the slip systems for nickel being active. Anisotropy was observed in nickel deposits, which was linked to texture evolution inside of the coating. Microsample testing guided the selection of 15mum layer of copper deposition followed by a 250 mum NDPA layer. Classical formulas for structural collapse were used to guide an experimental parametric study to establish a weight/volume efficient strut topology. Length, diameter and thickness were all investigated to determine the optimal column topology. The most optimal topology exists when Eulerian buckling, shell micro buckling and yielding failure modes all exist in a single geometric topology. Three macro-scale sandwich topologies (pyramidal, tetrahedral, and strut-reinforced-tetrahedral (SRT) were investigated with respect to strength-per-unit-weight. The

  10. Conceptual design of simplified PWR

    International Nuclear Information System (INIS)

    The authors believe the next generation nuclear power plant should be characterized by: (1) simplicity of design; (2) ease of operation and maintenance; (3) economic conformance with safety requirements; and (4) technologies easy to understand by the public. 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, the Japan Atomic Power Company (JAPC) supported by the other Japanese PWR Utilities, Electricite de France (EdF), Westinghouse (WH) and Mitsubishi Heavy Industry (MHI) have studied application of passive technologies at a power rating of about 1,000 MWe. The limited availability for location of nuclear power plant in Japan makes plants with higher power ratings more desirable. Using the AP-600 reference design as a basis, the authors enlarged the plant size to 3 -loops and added engineering features to conform with Japanese practice and Utilities' preference. The Simplified PWR (SPWR) program definitively confirmed the feasibility of a passive plant with an NSSS rating about 1,000 MWe and 3 loops

  11. Evaluation of trial design studies for an advanced marine reactor, (7)

    International Nuclear Information System (INIS)

    We have performed the numerical evaluation of shielding design for three advanced marine reactors (semi-integral PWR, integral PWR and self-pressurized PWR) under operational condition and hypothetical accident. Common calculational procedure and shielding material ingredient over the three reactors have been adopted for fair evaluation. (author)

  12. Manufacture of nuclear fuel elements for commercial PWR in China

    International Nuclear Information System (INIS)

    Yibin Nuclear Fuel Element Plant (YFP) under the leadership of China National Nuclear Corporation is sole manufacturer in China to specialize in the production of fuel assemblies and associated core components for commercial PWR nuclear power plant. At the early of 1980's, it began to manufacture fuel assemblies and associated core components for the first core of QINSHAN 300 MW nuclear power plant designed and built by China itself. With the development of nuclear power industry in China and the demand for localization of nuclear fuel elements in the early 1990's, YFP cooperated with FRAMATOME France in technology transfer for design and manufacturing of AFA 2G fuel assembly and successfully supplied the qualified fuel assemblies for the reloads of two units of GUANGDONG Da Ya Bay 900 MW nuclear power plant (Da Ya Bay NPP), and has achieved the localization of fuel assemblies and nuclear power plants. Meanwhile, it supplied fuel assemblies and associated core components for the first core and further reloads of Pakistan CHASHMA 300 MW nuclear power plant which was designed and built by China, and now it is manufacturing AFA 2G fuel assemblies and associated core components for the first core of two units of NPQJVC 600 MW nuclear power plant. From 2001 on, YFP will be able to supply Da Ya Bay NPP with the third generation of fuel assembly-AFA 3G which is to realize a strategy to develop the fuel assembly being of long cycle reload and high burn-up

  13. Experimental Results of the First Two Stages of an Advanced Transonic Core Compressor Under Isolated and Multi-Stage Conditions

    Science.gov (United States)

    Prahst, Patricia S.; Kulkarni, Sameer; Sohn, Ki H.

    2015-01-01

    NASA's Environmentally Responsible Aviation (ERA) Program calls for investigation of the technology barriers associated with improved fuel efficiency of large gas turbine engines. Under ERA the task for a High Pressure Ratio Core Technology program calls for a higher overall pressure ratio of 60 to 70. This mean that the HPC would have to almost double in pressure ratio and keep its high level of efficiency. The challenge is how to match the corrected mass flow rate of the front two supersonic high reaction and high corrected tip speed stages with a total pressure ratio of 3.5. NASA and GE teamed to address this challenge by using the initial geometry of an advanced GE compressor design to meet the requirements of the first 2 stages of the very high pressure ratio core compressor. The rig was configured to run as a 2 stage machine, with Strut and IGV, Rotor 1 and Stator 1 run as independent tests which were then followed by adding the second stage. The goal is to fully understand the stage performances under isolated and multi-stage conditions and fully understand any differences and provide a detailed aerodynamic data set for CFD validation. Full use was made of steady and unsteady measurement methods to isolate fluid dynamics loss source mechanisms due to interaction and endwalls. The paper will present the description of the compressor test article, its predicted performance and operability, and the experimental results for both the single stage and two stage configurations. We focus the detailed measurements on 97 and 100 of design speed at 3 vane setting angles.

  14. Exprimental Results of the First Two Stages of an Advanced Transonic Core Compressor Under Isolated and Multi-Stage Conditions.

    Science.gov (United States)

    Prahst, Patricia S.; Kulkarni, Sameer; Sohn, Ki H.

    2015-01-01

    NASA's Environmentally Responsible Aviation (ERA) Program calls for investigation of the technology barriers associated with improved fuel efficiency for large gas turbine engines. Under ERA, the highly loaded core compressor technology program attempts to realize the fuel burn reduction goal by increasing overall pressure ratio of the compressor to increase thermal efficiency of the engine. Study engines with overall pressure ratio of 60 to 70 are now being investigated. This means that the high pressure compressor would have to almost double in pressure ratio while keeping a high level of efficiency. NASA and GE teamed to address this challenge by testing the first two stages of an advanced GE compressor designed to meet the requirements of a very high pressure ratio core compressor. Previous test experience of a compressor which included these front two stages indicated a performance deficit relative to design intent. Therefore, the current rig was designed to run in 1-stage and 2-stage configurations in two separate tests to assess whether the bow shock of the second rotor interacting with the upstream stage contributed to the unpredicted performance deficit, or if the culprit was due to interaction of rotor 1 and stator 1. Thus, the goal was to fully understand the stage 1 performance under isolated and multi-stage conditions, and additionally to provide a detailed aerodynamic data set for CFD validation. Full use was made of steady and unsteady measurement methods to understand fluid dynamics loss source mechanisms due to rotor shock interaction and endwall losses. This paper will present the description of the compressor test article and its measured performance and operability, for both the single stage and two stage configurations. We focus the paper on measurements at 97% corrected speed with design intent vane setting angles.

  15. Glucitol-core containing gallotannins inhibit the formation of advanced glycation end-products mediated by their antioxidant potential.

    Science.gov (United States)

    Ma, Hang; Liu, Weixi; Frost, Leslie; Kirschenbaum, Louis J; Dain, Joel A; Seeram, Navindra P

    2016-05-18

    Glucitol-core containing gallotannins (GCGs) are polyphenols containing galloyl groups attached to a 1,5-anhydro-d-glucitol core, which is uncommon among naturally occurring plant gallotannins. GCGs have only been isolated from maple (Acer) species, including the red maple (Acer rubrum), a medicinal plant which along with the sugar maple (Acer saccharum), are the major sources of the natural sweetener, maple syrup. GCGs are reported to show antioxidant, α-glucosidase inhibitory, and antidiabetic effects, but their antiglycating potential is unknown. Herein, the inhibitory effects of five GCGs (containing 1-4 galloyls) on the formation of advanced glycation end-products (AGEs) were evaluated by MALDI-TOF mass spectroscopy, and BSA-fructose, and G.K. peptide-ribose assays. The GCGs showed superior activities compared to the synthetic antiglycating agent, aminoguanidine (IC50 15.8-151.3 vs. >300 μM) at the early, middle, and late stages of glycation. Circular dichroism data revealed that the GCGs were able to protect the secondary structure of BSA protein from glycation. The GCGs did not inhibit AGE formation by the trapping of reactive carbonyl species, namely, methylglyoxal, but showed free radical scavenging activities in the DPPH assay. The free radical quenching properties of the GCGs were further confirmed by electron paramagnetic resonance spectroscopy using ginnalin A (contains 2 galloyls) as a representative GCG. In addition, this GCG chelated ferrous iron, an oxidative catalyst of AGE formation, supported a potential antioxidant mechanism of antiglycating activity for these polyphenols. Therefore, GCGs should be further investigated for their antidiabetic potential given their antioxidant, α-glucosidase inhibitory, and antiglycating properties. PMID:27101975

  16. Atmea launches Atmea1 the mid-sized generation 3+ PWR you can rely on

    International Nuclear Information System (INIS)

    ATMEA, a daughter company of AREVA NP and Mitsubishi Heavy Industries, is developing and will supply ATMEA1, the most advanced 1100 MWe PWR plant with the combination of the unique set of competence and experience of its parent companies. This folder presents the ATMEA1 reactor main features. (J.S.)

  17. 压水堆核电站堆芯物理/热工水力耦合特性研究%Investigation on Coupling Characteristics of Neutronics/Thermal-hydraulics of PWR NPP Core

    Institute of Scientific and Technical Information of China (English)

    郑勇; 彭敏俊; 夏庚磊; 刘新凯

    2014-01-01

    采用RELAP5‐HD作为堆芯耦合计算程序,以秦山核电二期工程反应堆堆芯为研究对象,建立堆芯活性区的物理/热工水力耦合模型,在此基础上进行了稳态计算和掉棒事故仿真研究。结果表明,使用RELAP5‐HD计算得到的结果与电厂实测值符合较好,获得的掉棒事故参数曲线能准确反映事故工况下的参数变化趋势。稳态和事故工况的计算结果均符合堆芯物理/热工水力反馈效应的理论分析,证实了所建立的堆芯耦合模型的准确性,为下一步进行核电站系统的仿真分析提供基础。%In this paper ,an integrated neutronics/thermal‐hydraulic model for the reac‐tor of Qinshan Phase Ⅱ NPP project was developed ,using the RELAP5‐HD as core coupled computational code .Based on the coupled model ,the steady state calculation and the rod drop transient simulation were performed .The results show that the values obtained from RELAP5‐HD calculation agree well with the available measured data ,and the calculated accident curves can predict all major parameters trends of the transient with good accuracy .Both steady state and transient calculation results are in accordance with the theoretical analysis from the feedback aspect of coupled reactor neutronics/thermal‐hydraulics ,this demonstrates that a successful coupled model of Qinshan PhaseⅡ NPP core has been developed ,and the established model provides a good foundation for further simulation analysis of the nuclear power plant system .

  18. PWR operation and reloading: EDF experience and developments

    International Nuclear Information System (INIS)

    The large experience accumulated by EDF in PWR operation and reloading for about fifteen years required reliable and industrial techniques. Presently, about 54 units of 900 MWe and 1300 MWe PWR's are being operated through various fuel managements (three-batch cycle, four-batch cycle, plutonium recycling). EDF has developed two sets of automatized computational sequences with automatic generation of input data and core calculations for both, the Loading Pattern (LP) optimization and initialization of input data (fuel reshuffling), and for reload related calculations (safety evaluation, start-up physics tests prediction, operating data). As far as the LP search is concerned, it consists in a technique of 'trial and error' based upon knowledge and which is under very severe constraints. Then, reload values prediction and core following are performed with codes and calculational methods which have a high level of qualification and calibration over the large experience of in-core measurements. With respect to these different points, continuous efforts are done aimed at improving the overall reloading methods. Developments are being achieved at different levels. Because of load following perturbations, on-line and off-line core power distribution followings are evaluated with fast nodal CAROLINE code. This one is derived from the 3D design COCCINELLE code developed by EDF, and whose main features are 3D core calculations with optimized numerical schemes and fast resolution techniques, fuel thermal and neutronic feed-back effects modelling (pin by pin). As an alternative to LP manual design used currently, EDF has examined two possible approaches: expert system and optimization package. As far as automatic sequences are concerned, a new technique of automatic generation of input files was evaluated but priority has been given to improvements in physics by more 3D extensive calculations with the new COCCINELLE code

  19. PWR operation and reloading: EDF experience and developments

    International Nuclear Information System (INIS)

    The large experience accumulated by EDF in PWR operation and reloading for about fifteen years required reliable and industrial techniques. Presently, about 54 units of 900 MWe and 1300 MWe PWR's are being operated through various fuel management (three-batch cycle, four-batch cycle, plutonium recycling). EDF has developed two sets of automatized computational sequences with automatic generation of input data and core calculations for both, the Loading Pattern (LP) optimization and initialization of input data (fuel reshuffling), and for reload related calculations (safety evaluation, start-up physics tests prediction, operating data). As far as the LP search is concerned, it consists in a technique of ''trial and error' based upon knowledge and which is under very severe constraints. Then, reload values prediction and core following are performed with codes and calculational methods which have a high level of qualification and calibration over the large experience of in-core measurements. With respect to these different points, continuous efforts are done aimed at improving the overall reloading methods. Developments are being achieved at different levels. Because of load following perturbations, on-line and off-line core power distribution followings are evaluated with fast nodal CAROLINE code. This one is derived from the 3D design COCCINELLE code developed by EDF, and whose main features are 3D core calculations with optimized numerical schemes and fast resolution techniques, fuel thermal and neutronic feed-back effects modelling (pin by pin). As an alternative to LP manual design used currently, EDF has examined two possible approaches: Expert system and optimization package. As far as automatic sequences are concerned, a new technique of automatic generation of input files was evaluated but priority has been given to improvements in physics by more 3D extensive calculations with the new COCCINELLE code. (author). 4 refs, 3 figs

  20. Optimization of control area ventilation systems for Japanese PWR plants

    International Nuclear Information System (INIS)

    The nuclear power plant has been required to reduce the cost for the purpose of making the low-cost energy since several years ago in Japan. The Heating, Ventilating and Air Conditioning system in the nuclear power plant has been also required to reduce its cost. On the other hand the ventilation system should add the improvable function according to the advanced plant design. In response to these different requirements, the ventilation criteria and the design of the ventilation system have been evaluated and optimized in Japanese PWR Plant design. This paper presents the findings of the authors' study

  1. Modeling of PWR fuel at extended burnup

    International Nuclear Information System (INIS)

    Since FRAPCON-3 series was rolled out, many improvements have been implanted in fuel performance codes, based on most recent literature, to promote better predictions against current data. Much of this advances include: improving fuel gas release prediction, hydrogen pickup model, cladding corrosion, and many others. An example of those modifications has been new cladding materials has added into hydrogen pickup model to support M5™, ZIRLO™, and ZIRLO™ optimized family under pressurized water reactor (PWR) conditions. Recently some research have been made over USNRC's steady-state fuel performance code, assessments against FUMEX-III's data have concluded that FRAPCON provides best-estimate calculation of fuel performance. Face of this, a study is required to summarize all those modifications and new implementations, as well as to compare this result against FRAPCON's older version, scrutinizing FRAPCON-3 series documentation to understand the real goal and literature base of any improvements. We have concluded that FRAPCON's latest modifications are based on strong literature review. Those modifications were tested against most recent data to assure these results will be the best evaluation as possible. Many improvements have been made to allow USNRC to have an audit tool with the last improvements. (author)

  2. 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)

    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

  3. Assessment of radiological impact due to a hypothetical core disruptive accident for PFBR using an advanced atmospheric dispersion system

    International Nuclear Information System (INIS)

    Radiological impact due to air borne effluent dispersion from a hypothetical Core Disruptive Accident (CDA) scenario for Prototype Fast Breeder Reactor (PFBR) at Kalpakkam coastal site is estimated using an advanced system consisting of a 3-d meso-scale atmospheric model and a random walk particle dispersion model. A simulation of dispersion for CDA carried out for a typical summer day on 24th May 2003 predicted development of land-sea breeze circulation and Thermal Internal Boundary Layer (TIBL) at Kalpakkam site, which have been confirmed by observations. Analysis of dose distribution corresponding to predicted atmospheric conditions shows maximum dose from stack releases beyond the site boundary at about 4 km during TIBL fumigation and stable conditions respectively. A multi mode spatial concentration distribution has been noticed with diurnal meandering of wind under land sea breeze circulation. Over a meso-scale range of 25 km, turning of plume under sea breeze and maximum concentration along plume centerline at distances of 3 to 10 km have been noticed. The study has enabled to simulate the more complex meteorological situation that is actually present at the site. (author)

  4. Development of essential system technologies for advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Y. Y.; Hwang, Y. D.; Cho, B. H. and others

    1999-03-01

    Basic design of SMART adopts the new advanced technologies which were not applied in the existing 1000MWe PWR. However, the R and D experience on these advanced essential technologies is lacking in domestic nuclear industry. Recently, a research on these advanced technologies has been performed as a part of the mid-and-long term nuclear R and D program, but the research was limited only for the small scale fundamental study. The research on these essential technologies such as helically coiled tube steam generator, self pressurizer, core cooling by natural circulation required for the development of integral reactor SMART have not been conducted in full scale. This project, therefore, was performed for the development of analysis models and methodologies, system analysis and thermal hydraulic experiments on the essential technologies to be applied to the 300MWe capacity of integral reactor SMART and the advanced passive reactor expected to be developed in near future with the emphasis on experimental investigation. (author)

  5. Development of essential system technologies for advanced reactor

    International Nuclear Information System (INIS)

    Basic design of SMART adopts the new advanced technologies which were not applied in the existing 1000MWe PWR. However, the R and D experience on these advanced essential technologies is lacking in domestic nuclear industry. Recently, a research on these advanced technologies has been performed as a part of the mid-and-long term nuclear R and D program, but the research was limited only for the small scale fundamental study. The research on these essential technologies such as helically coiled tube steam generator, self pressurizer, core cooling by natural circulation required for the development of integral reactor SMART have not been conducted in full scale. This project, therefore, was performed for the development of analysis models and methodologies, system analysis and thermal hydraulic experiments on the essential technologies to be applied to the 300MWe capacity of integral reactor SMART and the advanced passive reactor expected to be developed in near future with the emphasis on experimental investigation. (author)

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

    International Nuclear Information System (INIS)

    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

  7. The synergy of PWR and PHWR in Korean nuclear power programme

    International Nuclear Information System (INIS)

    In Korea, 12 nuclear power plants are in commercial operation as of August 1997, and about 36% of the country's total electricity demand is provided by these nuclear power plants. Korea has adopted a two reactor PWR/PHWR (Candu) policy and this unique two reactor policy has received a lot of international attention. In general, many countries have adopted a two or multi reactor policy in order to both enhance the economic use of nuclear energy through various reactor technology developments and to stabilize the nuclear electricity generating system in view of safety. Korean experience has shown more synergy effects than mentioned above. First of all, the feed-back of technological advantages of PWR and PHWR has greatly contributed to the advancement of domestic nuclear industrial capabilities. The two reactor policy, PWR and PHWR, which are the most competitive commercial reactors available these days, has attracted with regard to the economic and operating advantages of the two reactors. In addition, the two reactor policy has contributed to the efficient use of spent PWR fuel and accordingly, the various options for the nuclear fuel cycle, recycling the spent PWR fuel into PHWR, such as CANFLEX-RU and DUPIC. (author)

  8. RELAP5 model for advanced neutron source reactor thermal-hydraulic transients, three-element-core design

    International Nuclear Information System (INIS)

    In order to utilize reduced enrichment fuel, the three-element-core design has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. However, the total flow rate through the core is greater and the pressure drop across the core is less so that the primary coolant pumps and heat exchangers are operating at a different point in their performance curves. This report describes the new RELAP5 input for the core components

  9. A neutronic study on advanced sodium cooled fast reactor cores with thorium blankets for effective burning of transuranic nuclides

    International Nuclear Information System (INIS)

    Highlights: • SFR burner core configurations are explored and analyzed for effective use of thorium blankets. • Thorium blankets can significantly improve SFR burner core performances. • No recycling or partial recycling of Th blankets with multi-batches is very effective. - Abstract: In this paper, new design concepts of sodium cooled fast reactor (SFR) cores having thorium blanket are suggested for pursuing effective burning of TRU (transuranics) nuclides from LWR spent fuels and their neutronic performances are analyzed. Several core configurations having different arrangements of thorium blankets are explored to improve the core performances and safety-related parameters including sodium void worth which is one of main concerns on safety of SFR cores. Specifically, axial and radial thorium blankets are considered for two type cores. The first one is the typical annular type cores having two different fuel regions where axial thorium blankets are placed in the axially central regions while the second one is the single fuel region cores having central non-fuel region where the axial blanket and radial blankets are considered. Also, the effects of the recycling options and fuel management schemes of the used thorium blanket on the core performances are analyzed. The core performance analyses show that thorium blankets with no recycling option and multi-batch fuel management schemes are very effective to improve the core performances including burnup reactivity swing, sodium void worth and TRU consumption rate

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

    International Nuclear Information System (INIS)

    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

  11. Preliminary Thermo-hydraulic Core Design Analysis of Korea Advanced Nuclear Thermal Engine Rocket for Space Application

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Seung Hyun; Lee, Jeong Ik; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    Nclear rockets improve the propellant efficiency more than twice compared to CRs and thus significantly reduce the propellant requirement. The superior efficiency of nuclear rockets is due to the combination of the huge energy density and a single low molecular weight propellant utilization. Nuclear Thermal Rockets (NTRs) are particularly suitable for manned missions to Mars because it satisfies a relatively high thrust as well as a high propellant efficiency. NTRs use thermal energy released from a nuclear fission reactor to heat a single low molecular weight propellant, i. e., Hydrogen (H{sub 2}) and then exhausted the extremely heated propellant through a thermodynamic nozzle to produce thrust. A propellant efficiency parameter of rocket engines is specific impulse (I{sub sp}) which represents the ratio of the thrust over the rate of propellant consumption. The difference of I{sub sp} makes over three times propellant savings of NTRs for a manned Mars mission compared to CRs. NTRs can also be configured to operate bimodally by converting the surplus nuclear energy to auxiliary electric power required for the operation of a spacecraft. Moreover, the concept and technology of NTRs are very simple, already proven, and safe. Thus, NTRs can be applied to various space missions such as solar system exploration, International Space Station (ISS) transport support, Near Earth Objects (NEOs) interception, etc. Nuclear propulsion is the most promising and viable option to achieve challenging deep space missions. Particularly, the attractions of a NTR include excellent thrust and propellant efficiency, bimodal capability, proven technology, and safe and reliable performance. The ROK has also begun the research for space nuclear systems as a volunteer of the international space race and a major world nuclear energy country. KANUTER is one of the advanced NTR engines currently under development at KAIST. This bimodal engine is operated in two modes of propulsion with 100 MW

  12. Preliminary Thermo-hydraulic Core Design Analysis of Korea Advanced Nuclear Thermal Engine Rocket for Space Application

    International Nuclear Information System (INIS)

    Nclear rockets improve the propellant efficiency more than twice compared to CRs and thus significantly reduce the propellant requirement. The superior efficiency of nuclear rockets is due to the combination of the huge energy density and a single low molecular weight propellant utilization. Nuclear Thermal Rockets (NTRs) are particularly suitable for manned missions to Mars because it satisfies a relatively high thrust as well as a high propellant efficiency. NTRs use thermal energy released from a nuclear fission reactor to heat a single low molecular weight propellant, i. e., Hydrogen (H2) and then exhausted the extremely heated propellant through a thermodynamic nozzle to produce thrust. A propellant efficiency parameter of rocket engines is specific impulse (Isp) which represents the ratio of the thrust over the rate of propellant consumption. The difference of Isp makes over three times propellant savings of NTRs for a manned Mars mission compared to CRs. NTRs can also be configured to operate bimodally by converting the surplus nuclear energy to auxiliary electric power required for the operation of a spacecraft. Moreover, the concept and technology of NTRs are very simple, already proven, and safe. Thus, NTRs can be applied to various space missions such as solar system exploration, International Space Station (ISS) transport support, Near Earth Objects (NEOs) interception, etc. Nuclear propulsion is the most promising and viable option to achieve challenging deep space missions. Particularly, the attractions of a NTR include excellent thrust and propellant efficiency, bimodal capability, proven technology, and safe and reliable performance. The ROK has also begun the research for space nuclear systems as a volunteer of the international space race and a major world nuclear energy country. KANUTER is one of the advanced NTR engines currently under development at KAIST. This bimodal engine is operated in two modes of propulsion with 100 MWth power and

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

    International Nuclear Information System (INIS)

    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.)

  14. Timing analysis of PWR fuel pin failures

    International Nuclear Information System (INIS)

    This report discusses research conducted to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B ampersand W) design (Oconee) and a Westinghouse (W) four-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin burnup, axial peaking factor, break size, emergency core cooling system availability, and main coolant pump trip on these times. The analysis was performed using the following codes: FRAPCON-2, for the calculation of steady-state fuel behavior; SCDAP/RELAP5/MOD3 and TRACPF1/MOD1, for the calculation of the transient thermal-hydraulic conditions in the reactor system; and FRAP-T6, for the calculation of transient fuel behavior. In addition to the calculation of fuel pin failure timing, this analysis provides a comparison of the predicted results of SCDAP/RELAP5/MOD3 and TRAC-PF1/MOD1 for large-break LOCA analysis. Using SCDAP/RELAP5/MOD3 thermal-hydraulic data, the shortest time intervals calculated between initiation of containment isolation and fuel pin failure are 10.4 seconds and 19.1 seconds for the B ampersand W and W plants, respectively. Using data generated by TRAC-PF1/MOD1, the shortest intervals are 10.3 seconds and 29.1 seconds for the B ampersand W and W plants, respectively. These intervals are for a double-ended, offset-shear, cold leg break, using the technical specification maximum peaking factor and applied to fuel with maximum design burnup. Using peaking factors commensurate with actual burnups would result in longer intervals for both reactor designs. This document provides appendices K and L of this report which provide plots for the timing analysis of PWR fuel pin failures for Oconee and Seabrook respectively

  15. Load-following operation of PWR plants

    International Nuclear Information System (INIS)

    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)

  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. Integral type small PWR with stand-alone safety

    Energy Technology Data Exchange (ETDEWEB)

    Makihara, Yoshiaki [Mitsubishi Heavy Industries, Ltd., Tokyo (Japan)

    2001-09-01

    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)

  18. MOX and UOX PWR fuel performances EDF operating experience

    International Nuclear Information System (INIS)

    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)

  19. Development of the evaluation methods in reactor safety analyses and core characteristics

    International Nuclear Information System (INIS)

    In order to support the safety reviews by NISA on reactor safety design, the computer codes are developed and maintained in the areas of safety analyses and core characteristics evaluation. In the code preparation of safety analyses, the TRACE code was prepared to conduct the safety analysis of LOCA for PWR. Also, the statistical safety evaluation method based on the three-dimensional neutron-kinetics/thermal-hydraulics coupled code SKETCH/TRACE has been prepared. In the core physics code preparation, the advanced neutron data library JENDL-4.0 were incorporated in the lattice physics code CASMO-4. The verification of improved CASMO-4 /SIMULATE-3 was continued by using core physics data. (author)

  20. Aspects of postoperative magnetic resonance imaging of patients with avascular necrosis of the femoral head, treated by advanced core decompression

    International Nuclear Information System (INIS)

    To analyze remodeling processes after advanced core decompression (ACD) in patients with avascular femoral head necrosis by means of 3T MRI and to identify indicators for clinical outcome considering the defect size and characteristics of the bone graft and of the neighboring regeneration tissue. Thirty-four hips, with preexisting preoperative MRIs in 21 cases, were examined 1-34 months (mean 12.7) postoperatively by 3T MRI. The volume of necrosis was measured manually pre- and postoperatively to calculate absolute as well as percentage necrosis reduction. The signal intensity of the bone graft was quantified using a 4-point scale. Border phenomena between the bone graft and bone were described and classified into groups. Wilcoxon sign-rank test was used to identify correlations between the analyzed items and clinical signs of femoral head collapse after a mean follow-up time of 28.6 months (10.4-46.8). Mean percentage reduction of necrosis was significantly higher in asymptomatic patients (59.36 %) compared to patients with signs of femoral head collapse (28.78 %, p = 0.008). Signal intensity of the bone graft increased in T1w and T2w TIRM sequences over time after surgery and was significantly higher in asymptomatic patients. Five border phenomena between the bone graft and healthy bone were identified. Among them, the so-called ''rail sign'' representing three layers of remodeling tissue correlated with the histological observations. A variety of border phenomena representing remodeling processes have been described using 3T MRI. Beneath the percentage amount of necrosis reduction, we identified the signal intensity of the bone graft as an indicator for clinical outcome. (orig.)

  1. Aspects of postoperative magnetic resonance imaging of patients with avascular necrosis of the femoral head, treated by advanced core decompression

    Energy Technology Data Exchange (ETDEWEB)

    Lazik, Andrea; Lauenstein, Thomas C.; Theysohn, Jens M. [University Hospital Essen, Department of Diagnostic and Interventional Radiology and Neuroradiology, Essen (Germany); Landgraeber, Stefan; Classen, Tim [University Hospital Essen, Department of Orthopedics, Essen (Germany); Kraff, Oliver [University of Duisburg-Essen, Erwin L. Hahn Institute for Magnetic Resonance Imaging, Essen (Germany)

    2015-10-15

    To analyze remodeling processes after advanced core decompression (ACD) in patients with avascular femoral head necrosis by means of 3T MRI and to identify indicators for clinical outcome considering the defect size and characteristics of the bone graft and of the neighboring regeneration tissue. Thirty-four hips, with preexisting preoperative MRIs in 21 cases, were examined 1-34 months (mean 12.7) postoperatively by 3T MRI. The volume of necrosis was measured manually pre- and postoperatively to calculate absolute as well as percentage necrosis reduction. The signal intensity of the bone graft was quantified using a 4-point scale. Border phenomena between the bone graft and bone were described and classified into groups. Wilcoxon sign-rank test was used to identify correlations between the analyzed items and clinical signs of femoral head collapse after a mean follow-up time of 28.6 months (10.4-46.8). Mean percentage reduction of necrosis was significantly higher in asymptomatic patients (59.36 %) compared to patients with signs of femoral head collapse (28.78 %, p = 0.008). Signal intensity of the bone graft increased in T1w and T2w TIRM sequences over time after surgery and was significantly higher in asymptomatic patients. Five border phenomena between the bone graft and healthy bone were identified. Among them, the so-called ''rail sign'' representing three layers of remodeling tissue correlated with the histological observations. A variety of border phenomena representing remodeling processes have been described using 3T MRI. Beneath the percentage amount of necrosis reduction, we identified the signal intensity of the bone graft as an indicator for clinical outcome. (orig.)

  2. A Consistent Comparative Study of Advanced Sodium-cooled Fast Burner Cores loaded with Thorium and Uranium-based Metallic Fuels

    International Nuclear Information System (INIS)

    We considered uranium-based metallic fuel of TRU-U-10Zr for driver fuel and thorium was considered as blanket because thorium blanket produces less amount of TRU than uranium blanket and use of thorium blanket leads to smaller sodium void worth than the use of uranium blanket due to the fact that the η-value increases much less with energy for 233U than for 239Pu and 232Th is less fissile than 238U. However, these cores using thorium blanket still have a large amount of TRU production from the driver fuels because the driver fuels contain a large amount of depleted uranium which leads to the production of TRU through neutron capture. The objective of this work is to consistently compare the neutronic performances of advanced sodium cooled fast reactor cores loaded with thorium and uraniumbased metallic fuels as driver fuel for TRU burning. Our main emphasis is given on the analyses of the differences in the core performance parameters. For consistent comparison, we used the same core configuration and all the same design parameters except for the fact that depleted uranium in uraniumbased fuel is replaced with thorium. We considered the cores having no thorium blanket and the cores having thorium blanket that were designed in our previous works

  3. MIC-SVM: Designing A Highly Efficient Support Vector Machine For Advanced Modern Multi-Core and Many-Core Architectures

    Energy Technology Data Exchange (ETDEWEB)

    You, Yang; Song, Shuaiwen; Fu, Haohuan; Marquez, Andres; Mehri Dehanavi, Maryam; Barker, Kevin J.; Cameron, Kirk; Randles, Amanda; Yang, Guangwen

    2014-08-16

    Support Vector Machine (SVM) has been widely used in data-mining and Big Data applications as modern commercial databases start to attach an increasing importance to the analytic capabilities. In recent years, SVM was adapted to the field of High Performance Computing for power/performance prediction, auto-tuning, and runtime scheduling. However, even at the risk of losing prediction accuracy due to insufficient runtime information, researchers can only afford to apply offline model training to avoid significant runtime training overhead. To address the challenges above, we designed and implemented MICSVM, a highly efficient parallel SVM for x86 based multi-core and many core architectures, such as the Intel Ivy Bridge CPUs and Intel Xeon Phi coprocessor (MIC).

  4. Study on severe accident mitigation measures for the development of PWR SAMG

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    In the development of the Severe Accident Management Guidelines (SAMG), it is very important to choose the main severe accident sequences and verify their mitigation measures. In this article, Loss-of-Coolant Accident (LOCA), Steam Generator Tube Rupture (SGTR), Station Blackout (SBO), and Anticipated Transients without Scram (ATWS) in PWR with 300 MWe are selected as the main severe accident sequences. The core damage progressions induced by the above-mentioned sequences are analyzed using SCDAP/RELAP5. To arrest the core damage progression and mitigate the consequences of severe accidents, the measures for the severe accident management (SAM) such as feed and bleed, and depressurizations are verified using the calculation. The results suggest that implementing feed and bleed and depressurization could be an effective way to arrest the severe accident sequences in PWR.

  5. Advances in methods of commercial FBR core characteristics analyses. Investigations of a treatment of the double-heterogeneity and a method to calculate homogenized control rod cross sections

    International Nuclear Information System (INIS)

    A standard data base for FBR core nuclear design is under development in order to improve the accuracy of FBR design calculation. As a part of the development, we investigated an improved treatment of double-heterogeneity and a method to calculate homogenized control rod cross sections in a commercial reactor geometry, for the betterment of the analytical accuracy of commercial FBR core characteristics. As an improvement in the treatment of double-heterogeneity, we derived a new method (the direct method) and compared both this and conventional methods with continuous energy Monte-Carlo calculations. In addition, we investigated the applicability of the reaction rate ratio preservation method as a advanced method to calculate homogenized control rod cross sections. The present studies gave the following information: (1) An improved treatment of double-heterogeneity: for criticality the conventional method showed good agreement with Monte-Carlo result within one sigma standard deviation; the direct method was consistent with conventional one. Preliminary evaluation of effects in core characteristics other than criticality showed that the effect of sodium void reactivity (coolant reactivity) due to the double-heterogeneity was large. (2) An advanced method to calculate homogenize control rod cross sections: for control rod worths the reaction rate ratio preservation method agreed with those produced by the calculations with the control rod heterogeneity included in the core geometry; in Monju control rod worth analysis, the present method overestimated control rod worths by 1 to 2% compared with the conventional method, but these differences were caused by more accurate model in the present method and it is considered that this method is more reliable than the conventional one. These two methods investigated in this study can be directly applied to core characteristics other than criticality or control rod worth. Thus it is concluded that these methods will

  6. Advanced computational methods for the assessment of reactor core behaviour during reactivity initiated accidents. Final report; Fortschrittliche Rechenmethoden zum Kernverhalten bei Reaktivitaetsstoerfaellen. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Pautz, A.; Perin, Y.; Pasichnyk, I.; Velkov, K.; Zwermann, W.; Seubert, A.; Klein, M.; Gallner, L.; Krzycacz-Hausmann, B.

    2012-05-15

    The document at hand serves as the final report for the reactor safety research project RS1183 ''Advanced Computational Methods for the Assessment of Reactor Core Behavior During Reactivity-Initiated Accidents''. The work performed in the framework of this project was dedicated to the development, validation and application of advanced computational methods for the simulation of transients and accidents of nuclear installations. These simulation tools describe in particular the behavior of the reactor core (with respect to neutronics, thermal-hydraulics and thermal mechanics) at a very high level of detail. The overall goal of this project was the deployment of a modern nuclear computational chain which provides, besides advanced 3D tools for coupled neutronics/ thermal-hydraulics full core calculations, also appropriate tools for the generation of multi-group cross sections and Monte Carlo models for the verification of the individual calculational steps. This computational chain shall primarily be deployed for light water reactors (LWR), but should beyond that also be applicable for innovative reactor concepts. Thus, validation on computational benchmarks and critical experiments was of paramount importance. Finally, appropriate methods for uncertainty and sensitivity analysis were to be integrated into the computational framework, in order to assess and quantify the uncertainties due to insufficient knowledge of data, as well as due to methodological aspects.

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

    International Nuclear Information System (INIS)

    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

  8. Studies on influence of sodium void reactivity effect on the concept of the core and safety of advanced fast reactor

    International Nuclear Information System (INIS)

    The paper is devoted to studies on influence of sodium void reactivity effect (SVRE) on safety and technical and economical characteristics of BN-1200 type reactor. Different core options are considered as applied to this reactor. These core options differ in designs, dimensions and, hence, SVRE value. It is shown by the analysis that most flattened core with sodium plenum at the top assures reactor self-protection under beyond design basis accident conditions. Sodium plenum abandonment and core height increase causing SVRE increase deteriorate reactor self-protection, but at the same time, improve some technical and economical characteristics of the reactor. Issues of choosing optimal core design under these conditions are discussed. (author)

  9. STUDY OF THE THERMAL STRATIFICATION IN PWR REACTORS AND THE PTS (PRESSURIZED THERMAL SHOCK) PHENOMENON

    OpenAIRE

    ROMERO HAMERS, ADOLFO

    2014-01-01

    In the event of hypothetical accident scenarios in PWR, emergency strategies have to be mapped out, in order to guarantee the reliable removal of decay heat from the reactor core, also in case of component breakdown. One essential passive heat removal mechanism is the reflux condensation cooling mode. This mode can appear for instance during a small break loss-of-coolant-accident (LOCA) or because of loss of residual heat removal (RHR) system during mid loop operation at plant outage after th...

  10. Neutron leakage treatment in reactor physics. Consequences for predicting core characteristics

    International Nuclear Information System (INIS)

    New generations of simulation tools responding to the challenges brought by the advanced features of both 3rd+ generation Pressurized Water Reactor (PWR) cores and 4th generation sodium fast neutron reactor (SFR) cores are taking shape. The developments of new simulations tools are also motivated by strict requirements of nuclear safety authorities. The new tools have the objective of setting new reference standards for neutronic prediction and will take advantage of innovative algorithms which have been implemented in existing CEA codes, such as ERANOS (fast reactors) and APOLLO2 (PWR); the new codes should at the same time remove remaining calculation errors. Although innovative algorithms have been filling the gaps which did exist 40 years ago between tools specifically dedicated to either thermal neutron cores or fast neutron ones, there remains a series of algorithms which deserve particular attention: the treatment of leakage in cell calculations. This paper describes methods for treating neutron leakage in self-shielding calculations with the sub-group method, and in the cell balance calculation. Applications of the MOC method of solution to treat neutron leakage are described. The application of the MOC can eliminate approximations at the cell interfaces while maintaining precise neutron leakage treatment. The new APOLLO3® code, presently under development at CEA, is candidate for hosting such algorithms. (author)

  11. A Novel Burnable Absorber Concept for PWR: BigT (Burnable Absorber-Integrated Guide Thimble)

    International Nuclear Information System (INIS)

    This paper presents the essential BigT design concepts and its lattice neutronic characteristics. Neutronic performance of a newly-proposed BA concept for PWR named BigT is investigated in this study. Preliminary lattice analyses of the BigT absorber-loaded WH 17x17 fuel assembly show a high potential of the concept as it performs relatively well in comparison with commercial burnable absorber technologies, especially in managing reactivity depletion and peaking factor. A sufficiently high control rod worth can still be obtained with the BigT absorbers in place. It is expected that with such performance and design flexibilities, any loading pattern and core management objective, including a soluble boron-free PWR, can potentially be fulfilled with the BigT absorbers. Future study involving full 3D reactor core simulations with the BigT absorbers shall hopefully verify this hypothesis. A new burnable absorber design for Pressurized Water Reactor (PWR) named 'Burnable absorber-Integrated control rod Guide Thimble' (BigT) was recently proposed. Unlike conventional burnable absorber (BA) technologies, the BigT integrates BA materials directly into the guide thimble but still allows insertion of control rod (CR). In addition, the BigT offers a variety of design flexibilities such that any loading pattern and core management objective can potentially be fulfilled

  12. A Novel Burnable Absorber Concept for PWR: BigT (Burnable Absorber-Integrated Guide Thimble)

    Energy Technology Data Exchange (ETDEWEB)

    Yahya, Mohdsyukri; Kim, Yonghee [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Chung, Chang Kyu [KEPCO Engineering and Construction Company, Daejeon (Korea, Republic of)

    2014-05-15

    This paper presents the essential BigT design concepts and its lattice neutronic characteristics. Neutronic performance of a newly-proposed BA concept for PWR named BigT is investigated in this study. Preliminary lattice analyses of the BigT absorber-loaded WH 17x17 fuel assembly show a high potential of the concept as it performs relatively well in comparison with commercial burnable absorber technologies, especially in managing reactivity depletion and peaking factor. A sufficiently high control rod worth can still be obtained with the BigT absorbers in place. It is expected that with such performance and design flexibilities, any loading pattern and core management objective, including a soluble boron-free PWR, can potentially be fulfilled with the BigT absorbers. Future study involving full 3D reactor core simulations with the BigT absorbers shall hopefully verify this hypothesis. A new burnable absorber design for Pressurized Water Reactor (PWR) named 'Burnable absorber-Integrated control rod Guide Thimble' (BigT) was recently proposed. Unlike conventional burnable absorber (BA) technologies, the BigT integrates BA materials directly into the guide thimble but still allows insertion of control rod (CR). In addition, the BigT offers a variety of design flexibilities such that any loading pattern and core management objective can potentially be fulfilled.

  13. Proceeding of JSPS-CAS Core University Program seminar on production and control of high performance plasmas with advanced plasma heating and diagnostic systems

    International Nuclear Information System (INIS)

    The JSPS-CAS Core University Program (CUP) seminar on 'Production and control of high performance plasmas with advanced plasma heating and diagnostic systems' took place in Guilin Bravo Hotel, Guilin, China, 1-4 November 2010. This seminar was organized in the framework of CUP in the field of plasma and nuclear fusion. Two special talks and 46 oral talks were presented in the seminar including 36 Chinese, 18 Japanese and 4 Korean attendees. Production and control of high performance plasmas is a crucial issue for realizing an advanced nuclear fusion reactor in addition to developments of advanced plasma heating and diagnostics. This seminar was motivated along the issues. Results in the field of fusion experiments obtained through CUP activities during recent two years were summarized. Possible direction of future collaboration and further encouragement of scientific activity of younger scientists were also discussed in this seminar with future experimental plans in both countries. (author)

  14. Proceeding of JSPS-CAS core university program seminar on production and control of high performance plasmas with advanced plasma heating and diagnostic systems

    International Nuclear Information System (INIS)

    The JSPS-CAS Core University Program (CUP) seminar on 'Production and control of high performance plasmas with advanced plasma heating and diagnostic systems' took place in Shiner hotel, Lijiang, China, 4-7 November 2008. This seminar was organized in the framework of CUP in the field of plasma and nuclear fusion. One special talk and 34 oral talks were presented in the seminar including 16 Japanese attendees. Production and control of high performance plasmas is a crucial issue for realizing an advanced nuclear fusion reactor in addition to developments of advanced plasma heating and diagnostics. This seminar was motivated along the issues. Results obtained from CUP activities during recent four years were summarized. Several crucial issues to be resolved near future were also extracted in this seminar. The 31 of the papers are indexed individually. (J.P.N.)

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

    International Nuclear Information System (INIS)

    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

  16. Thermal-hydraulic model verification calculation of PWR tests

    International Nuclear Information System (INIS)

    Test PWR 5 determined forces and pressure differences across the reactor pressure vessel and the internals in the test vessel during the first 80 ms by means of the present test data tapes. Furthermore a comparison between the measured data and those determined with the aid of the LECK program system is carried out. The following results were obtained in this connection: The qualitative pattern as compared between calculation and measurement shows a good agreement. Higher pressure differences resulted across the components due to the higher pressure gradients in the initial phase of the blowdown verification in the calculations. The best agreement of the pressure gradients was obtained with the verification calculations for a rupture opening time of 6 ms. Since there was no fluid/structural-dynamic coupling it was not possible to simulate the premature pressure reduction within the core barrel. The distribution of the initial temperature in the calculation did not always agree with that during the test. (orig.)

  17. 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.

  18. Evaluation of fire probabilistic safety assessment for a PWR plant

    International Nuclear Information System (INIS)

    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)

  19. Conceptual design of SPWR, a PWR with enhanced passive safety

    International Nuclear Information System (INIS)

    A conceptual design has been carried out on a new type of integrated pressurized water reactor, SPWR (System-integrated PWR). This reactor installs a poison tank (borated water filled) in the reactor vessel instead of control rod drive system. Three hydraulic pressure valves are installed as the upper interface between the poison tank and primary coolant. A 700MWe power plant with twin 1100MWt SPWRs which are installed in a reactor building has been studied. Design and analysis have been made on the reactor core, reactor (reactor vessel, steam generator, main circulating pump, pressurizer, poison tank and their integration), plant systems (main and sub systems), layout, construction scheme, operation and maintenance, safety related components, reactor dynamics, economics and R and D needs. Passive safety features are also studied. (author)

  20. Activity transport models for PWR primary circuits

    International Nuclear Information System (INIS)

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

  1. Program of monitoring PWR fuel in Spain

    International Nuclear Information System (INIS)

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

  2. Studies on influence of sodium void reactivity effect on the concept of the core and safety of advanced fast reactor

    International Nuclear Information System (INIS)

    The paper is devoted to studies on the influence of the sodium void reactivity effect (SVRE) on the safety and technical and economical characteristics of the BN-1200-type reactor. Different core options are considered for application to this reactor. These core options differ in design, dimensions, and, hence, SVRE value. It is shown by the analysis that the most flattened core with sodium plenum at the top assures reactor self-protection under beyond-design-basis accident conditions. Sodium plenum abandonment and core height increase causing an SVRE value increase deteriorate reactor self-protection, but at the same time, improve some technical and economical characteristics of the reactor. Self-protection means the possibility to avoid rapid core meltdown under conditions of the above-listed beyond-design accidents. The possibility of controlling beyond-design accidents (for instance, by restoring the power supply of the main pumps in a rather short time) is taken into account. Issues of choosing the optimal core design under these conditions are discussed. (author)

  3. PWR control rod ejection analysis with the numerical nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hursin, M.; Kochunas, B.; Downar, T. J. [Univ. of California at Berkeley, Berkeley (Canada)

    2008-10-15

    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

  4. PWR control rod ejection analysis with the numerical nuclear reactor

    International Nuclear Information System (INIS)

    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

  5. Limitations of detecting inadequate core cooling with core exit thermocouples

    International Nuclear Information System (INIS)

    USNRC has suggested that the thermocouples (TCs) currently installed at the flow exit of a PWR core could be used to detect a condition of inadequate core cooling (ICC). The use of these TCs has been assumed in the USNRC Regulatory Guide 1.97. PWR vendors have responded to this guideline by proposing ICC instrumentation and procedure packages that include the use of core-exit TCs as a principal means of ICC detection. The core-exit TCs are judged to be able to detect an ICC condition because steam in the core will be superheated by the fuel rods and then flow past the TCs during an accident. The detection of superheat in the fluid stream constitutes the indirect detection of a core uncovery and heatup, or ICC. Data have been analyzed from four experiments conducted in the Loss-of-Fluid Test (LOFT) facility and the results indicate that there are two limitations to the detection of ICC by core exit TCs that should be resolved before reliance can be placed in the measurement. The LOFT TCs are described and these limitations are discussed in this paper

  6. New genetic algorithms (GA) to optimize PWR reactors

    International Nuclear Information System (INIS)

    The Haling Power Distribution (HPD) has been applied in a unique process to greatly accelerate the in-core fuel management optimization calculations. These calculations involve; the arrangement of fuel assemblies (FAs) and the placement of Burnable Poisons (BPs) in the fresh FAs. The HPD deals only with the arrangement of FAs. The purpose of this paper is to describe past uses of the HPD, provide an example selected from many similar calculations to explain why and how it can be used, and also to show its effectiveness as a filter in the GARCO GA code. The GARCO (Genetic Algorithm Reactor Core Optimization) is an innovative GA code that was developed by modifying the classical representation of the genotype and GA operators. A reactor physics code evaluates the LPs in the population using the HPD Method, which rapidly depletes the core in a single depletion step with a constant power distribution. The HPD is used basically in GARCO as a filter to eliminate invalid LPs created by the genetic operators, to choose a reference LP for BP optimization, and to create an initial population for simultaneous optimization of the LP and BP placement into the core. The accurate depletion calculation of the LP with BPs is done with the coupled lattice and reactor physics CASMO-4/SIMULATE3 package. However, the fact that these codes validate safety of the core with the added BP placement design also validates the use of the HPD method. The calculations are applied to the TMI-1 core as an example PWR providing concrete results

  7. Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel

    International Nuclear Information System (INIS)

    A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO2 fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

  8. Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel

    Energy Technology Data Exchange (ETDEWEB)

    Refeat, Riham Mahmoud [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.

    2015-12-15

    A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO{sub 2} fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.

  9. Recent advances in the synthesis of Fe{sub 3}O{sub 4}@AU core/shell nanoparticles

    Energy Technology Data Exchange (ETDEWEB)

    Salihov, Sergei V. [National University of Science and Technology MISiS, Leninskiy, Building 9, Moscow, 119049, Russian Federation, (Russian Federation); Ivanenkov, Yan A.; Krechetov, Sergei P.; Veselov, Mark S. [Moscow Institute of Physics and Technology (State University), 9 Institutskiy lane, Dolgoprudny City, Moscow Region, 141700 (Russian Federation); Sviridenkova, Natalia V.; Savchenko, Alexander G. [National University of Science and Technology MISiS, Leninskiy, Building 9, Moscow, 119049, Russian Federation, (Russian Federation); Klyachko, Natalya L. [National University of Science and Technology MISiS, Leninskiy, Building 9, Moscow, 119049, Russian Federation, (Russian Federation); Moscow State University, Chemistry Department, Lenins kie gory, Building 1/3, GSP-1, Moscow, 119991 (Russian Federation); Golovin, Yury I. [Moscow State University, Chemistry Department, Lenins kie gory, Building 1/3, GSP-1, Moscow, 119991 (Russian Federation); Chufarova, Nina V., E-mail: chnv@pharmcluster.ru [Moscow Institute of Physics and Technology (State University), 9 Institutskiy lane, Dolgoprudny City, Moscow Region, 141700 (Russian Federation); Beloglazkina, Elena K. [National University of Science and Technology MISiS, Leninskiy, Building 9, Moscow, 119049, Russian Federation, (Russian Federation); Moscow State University, Chemistry Department, Lenins kie gory, Building 1/3, GSP-1, Moscow, 119991 (Russian Federation); Majouga, Alexander G., E-mail: majouga@org.chem.msu.ru [National University of Science and Technology MISiS, Leninskiy, Building 9, Moscow, 119049, Russian Federation, (Russian Federation); Moscow State University, Chemistry Department, Lenins kie gory, Building 1/3, GSP-1, Moscow, 119991 (Russian Federation)

    2015-11-15

    Fe{sub 3}O{sub 4}@Au core/shell nanoparticles have unique magnetic and optical properties. These nanoparticles are used for biomedical applications, such as magnetic resonance imaging, photothermal therapy, controlled drug delivery, protein separation, biosensors, DNA detection, and immunosensors. In this review, recent methods for the synthesis of core/shell nanoparticles are discussed. We divided all of the synthetic methods in two groups: methods of synthesis of bi-layer structures and methods of synthesis of multilayer composite structures. The latter methods have a layer of “glue” material between the core and the shell. - Highlights: • Fe{sub 3}O{sub 4} nanoparticles are promising for biomedical applications but have some disadvantages. • Covering Fe{sub 3}O{sub 4} nanoparticles with Au shell leads to better stability and biocompatibility. • Core/shell nanoparticles are widely used for biomedical applications. • There are two types of Fe{sub 3}O{sub 4}@Au core/shell nanoparticles structures: bi-layer and multilayer composite. • Different synthetic methods enable production of nanoparticles of different sizes.

  10. Simulation model of a PWR power plant

    International Nuclear Information System (INIS)

    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)

  11. 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)

  12. Effect of coolant chemistry on PWR radiation transport processes

    International Nuclear Information System (INIS)

    The effect of various PWR-type coolant chemistry regimes on the behavior of corrosion products has been studied in the DIDO Water Loop at Harwell. While the extent of in-core spinel deposition is influenced by pH in a manner to be expected from the temperature coefficient of solubility of nickel-iron spinel, there is evidence that boric acid plays a role apart from its influence on pH. Out-of-core deposition of active cobalt on stainless steel takes place largely in the chromium-rich inner oxide layer, and there is also significant uptake of corrosion products into the film on Zircaloy. Deposition depends on flow characteristics in different ways for different elements. The evidence suggests that in DWL soluble species are dominant in out-of-core deposition processes for corrosion products. The adsorption of cobalt in zirconium oxide provides a route for deposition on fuel elements which may in some circumstances be more significant than spinel deposition. A most important factor governing deposition behavior is surface condition; the influence of weld regions and the effect of varying pretreatment conditions have both been demonstrated. Some alternative chemistry regimes have been explored, but do not appear to offer any advantages with respect to activity transport control over the more conventional regime based on lithium hydroxide and hydrogen dosing. 8 refs., 26 figs., 28 tabs

  13. PWR loading pattern optimization using Harmony Search algorithm

    International Nuclear Information System (INIS)

    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

  14. The three-dimensional PWR transient code ANTI; rod ejection test calculation

    International Nuclear Information System (INIS)

    ANTI is a computer program being developed for three-dimensional coupled neutronics and thermal-hydraulics description of a PWR core under transient conditions. In this report a test example calculated by the program is described. The test example is a simulation of a control rod ejection from a very small reactor core (to save somputing time). In order to show the influence of cross flow between adjacent fuel elements the same calculation was performed both with the cross flow option and with closed hydraulic channels. (author)

  15. Simulation of a low-pressure severe accident scenario in a PWR with ATHLET-CD

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, Mathias; Koch, Marco K. [Bochum Univ. (Germany). Reactor Simulation and Safety Group

    2013-07-01

    The plant behavior of a Pressurized Water Reactor (PWR) during a severe accident scenario is analyzed with system code ATHLET-CD Mod. 2.2C in order to assess the code capabilities in terms of the late-phase of the core degradation. For this purpose a severe accident sequence caused by a Station Black-out and a large break in the primary cooling system is simulated both without any accident management measures and with a delayed reflooding of the substantially degraded core. Selected code results are presented in this paper. (orig.)

  16. Development of high temperature adsorbent in PWR primary system

    International Nuclear Information System (INIS)

    Radiation exposure reduction in PWR is one of the most important problems to be solved. We have developed a high temperature Co adsorbent (HTA), which could be directly applied under primary reactor coolant conditions. This adsorbent was Fe-Ti-O system ceramics, and was fabricated to a suitable form for using in a packed column. Through those experiments of adsorption tests, compatibility tests, leaching tests and hot loop tests, it was found that HTA had superior adsorption capability to not only Co and Ni-ion but also many other transition metal ions. And it was also found that HTA was compatible with high temperature water, as well as advantageous for its waste solidification. Based on the experimental results, dose reduction effect was evaluated by a computer code. From this evaluation, it was found that more than 50 % dose reduction could be expected, when an advanced reactor coolant clean-up (RCC) system with HTA would be realized. (author)

  17. ANDREA: Advanced nodal diffusion code for reactor analysis

    International Nuclear Information System (INIS)

    A new macro code is being developed at NRI which will allow coupling of the advanced thermal-hydraulics model with neutronics calculations as well as efficient use in core loading pattern optimization process. This paper describes the current stage of the macro code development. The core simulator is based on the nodal expansion method, Helios lattice code is used for few group libraries preparation. Standard features such as pin wise power reconstruction and feedback iterations on critical control rod position, boron concentration and reactor power are implemented. A special attention is paid to the system and code modularity in order to enable flexible and easy implementation of new features in future. Precision of the methods used in the macro code has been verified on available benchmarks. Testing against Temelin PWR operational data is under way (Authors)

  18. Methodology for the LABIHS PWR simulator modernization

    Energy Technology Data Exchange (ETDEWEB)

    Jaime, Guilherme D.G.; Oliveira, Mauro V., E-mail: gdjaime@ien.gov.b, E-mail: mvitor@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    The Human-System Interface Laboratory (LABIHS) simulator is composed by a set of advanced hardware and software components whose goal is to simulate the main characteristics of a Pressured Water Reactor (PWR). This simulator serves for a set of purposes, such as: control room modernization projects; designing of operator aiding systems; providing technological expertise for graphical user interfaces (GUIs) designing; control rooms and interfaces evaluations considering both ergonomics and human factors aspects; interaction analysis between operators and the various systems operated by them; and human reliability analysis in scenarios considering simulated accidents and normal operation. The simulator runs in a PA-RISC architecture server (HPC3700), developed nearby 2000's, using the HP-UX operating system. All mathematical modeling components were written using the HP Fortran-77 programming language with a shared memory to exchange data from/to all simulator modules. Although this hardware/software framework has been discontinued in 2008, with costumer support ceasing in 2013, it is still used to run and operate the simulator. Due to the fact that the simulator is based on an obsolete and proprietary appliance, the laboratory is subject to efficiency and availability issues, such as: downtime caused by hardware failures; inability to run experiments on modern and well known architectures; and lack of choice of running multiple simulation instances simultaneously. This way, there is a need for a proposal and implementation of solutions so that: the simulator can be ported to the Linux operating system, running on the x86 instruction set architecture (i.e. personal computers); we can simultaneously run multiple instances of the simulator; and the operator terminals run remotely. This paper deals with the design stage of the simulator modernization, in which it is performed a thorough inspection of the hardware and software currently in operation. Our goal is to

  19. Methodology for the LABIHS PWR simulator modernization

    International Nuclear Information System (INIS)

    The Human-System Interface Laboratory (LABIHS) simulator is composed by a set of advanced hardware and software components whose goal is to simulate the main characteristics of a Pressured Water Reactor (PWR). This simulator serves for a set of purposes, such as: control room modernization projects; designing of operator aiding systems; providing technological expertise for graphical user interfaces (GUIs) designing; control rooms and interfaces evaluations considering both ergonomics and human factors aspects; interaction analysis between operators and the various systems operated by them; and human reliability analysis in scenarios considering simulated accidents and normal operation. The simulator runs in a PA-RISC architecture server (HPC3700), developed nearby 2000's, using the HP-UX operating system. All mathematical modeling components were written using the HP Fortran-77 programming language with a shared memory to exchange data from/to all simulator modules. Although this hardware/software framework has been discontinued in 2008, with costumer support ceasing in 2013, it is still used to run and operate the simulator. Due to the fact that the simulator is based on an obsolete and proprietary appliance, the laboratory is subject to efficiency and availability issues, such as: downtime caused by hardware failures; inability to run experiments on modern and well known architectures; and lack of choice of running multiple simulation instances simultaneously. This way, there is a need for a proposal and implementation of solutions so that: the simulator can be ported to the Linux operating system, running on the x86 instruction set architecture (i.e. personal computers); we can simultaneously run multiple instances of the simulator; and the operator terminals run remotely. This paper deals with the design stage of the simulator modernization, in which it is performed a thorough inspection of the hardware and software currently in operation. Our goal is to

  20. Advanced methods in teaching reactor physics

    International Nuclear Information System (INIS)

    Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for (nuclear power plant) operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software.

  1. Optimal implantation depth and adherence to guidelines on permanent pacing to improve the results of transcatheter aortic valve replacement with the medtronic corevalve system: The CoreValve prospective, international, post-market ADVANCE-II study

    NARCIS (Netherlands)

    A.S. Petronio (Anna S.); J.-M. Sinning (Jan-Malte); N.M. van Mieghem (Nicolas); G. Zucchelli (Giulio); G. Nickenig (Georg); R. Bekeredjian (Raffi); B. Bosmans; F. Bedogni (Francesco); M. Branny (Marian); K. Stangl (Karl); J. Kovac (Jan); M. Schiltgen (Molly); S. Kraus (Stacia); P.P.T. de Jaegere (Peter)

    2015-01-01

    textabstractObjectives The aim of the CoreValve prospective, international, post-market ADVANCE-II study was to define the rates of conduction disturbances and permanent pacemaker implantation (PPI) after transcatheter aortic valve replacement with the Medtronic CoreValve System (Minneapolis, Minnes

  2. Neutronics methods, models, and applications at the Idaho National Engineering Laboratory for the advanced neutron source reactor three-element core design

    International Nuclear Information System (INIS)

    A summary of the methods and models used to perform neutronics analyses on the Advanced Neutron Source reactor three-element core design is presented. The applications of the neutral particle Monte Carlo code MCNP are detailed, as well as the expansion of the static role of MCNP to analysis of fuel cycle depletion calculations. Results to date of these applications are presented also. A summary of the calculations not yet performed is also given to provide a open-quotes to-doclose quotes list if the project is resurrected

  3. A new type of simulator, developped for Super-Phenix breeder and for PWR, allows phenomena analysis as well as process control training

    International Nuclear Information System (INIS)

    After an outline of the characteristics of study-type and training-type simulators, a new dual-purpose system is described which has been developed by the French CISI company, for a much lower cost than training simulator's one. For Super-Phenix reactor, the model describes the core, the primary vessel, reactor kinetics, the intermediate heat exchanger, the steam generator, piping and secondary tank, the turbine. For PWR, following components are modellized: core, steam generators, pressurizer, boiler. The control room includes two benchboards (trainee and instructor). A module design allows to move quickly from Super-Phenix configuration to PWR one

  4. Improvements to PWR type reactors

    International Nuclear Information System (INIS)

    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

  5. A review of the DOE/ARSAP core melt progression program supporting design certification for advanced light water reactors

    International Nuclear Information System (INIS)

    An important element of the safety approach for AP600 is in-vessel retention (IVR). System design features are provided which enable the reactor cavity to be flooded in the event of a core melt accident. The ARSAP is performing work demonstrating that core melt materials would be retained and cooled within the vessel by means of this cavity flooding, and that ex-vessel threats to the containment are thereby avoided. This paper address in-vessel melt progression analyses being performed by ARSAP which are providing input to IVR-related issues such as in-vessel steam explosion and melt stream impingement/vessel wall ablation (the issue of melt coolability in the vessel lower head, by virtue of external water cooling, is not sensitive to details of core melt progression and has been handled separately). This paper also reviews the in-vessel core melt progression database as well as results from the TMI-2 accident which are the foundation of the current analyses for AP600. (author). 15 refs

  6. Design for reactor core safety in nuclear power plants

    International Nuclear Information System (INIS)

    This Guide covers the neutronic, thermal, hydraulic, mechanical, chemical and irradiation considerations important to the safe design of a nuclear reactor core. The Guide applies to the types of thermal neutron reactor power plants that are now in common use and fuelled with oxide fuels: advanced gas cooled reactor (AGR), boiling water reactor (BWR), pressurized heavy water reactor (PHWR) (pressure tube and pressure vessel type) and pressurized water reactor (PWR). It deals with the individual components and systems that make up the core and associated equipment and with design provisions for the safe operation of the core and safe handling of the fuel and other core components. The Guide discusses the reactor vessel internals and the reactivity control and shutdown devices mounted on the vessel. Possible effects on requirements for the reactor coolant, the reactor coolant system and its pressure boundary (including the pressure vessel) are considered only as far as necessary to clarify the interface with the Safety Guide on Reactor Coolant and Associated Systems in Nuclear Power Plants (IAEA Safety Series No. 50-SG-D13) and other Guides. In relation to instrumentation and control systems the guidance is mainly limited to functional requirements

  7. Long term coolability of a core melt

    International Nuclear Information System (INIS)

    One of the problems which must be solved in severe accidents is the melt concrete interaction which occurs when the core debris penetrates the lower vessel head and contacts the basement. To prevent these consequences a core catcher concept is considered to be integrated into a new PWR design based on the standard German PWR. The core catcher achieves coolability by spreading and fragmentation of the ex-vessel core melt based on the process of water inlet from the bottom through the melt. In order to identify the dominant processes of flooding the melt from the bottom experiments in laboratory scale have been carried out. To get more detailed information on the very important process of water penetration into the melt, a simulant experiment has been conducted using a transparent plastic melt with the typical viscosity behaviour of an oxidic corium melt and a temperature allowing evaporation of water. (orig.(DG)

  8. Optimization of small long-life PWR based on thorium fuel

    Science.gov (United States)

    Subkhi, Moh Nurul; Suud, Zaki; Waris, Abdul; Permana, Sidik

    2015-09-01

    A conceptual design of small long-life Pressurized Water Reactor (PWR) using thorium fuel has been investigated in neutronic aspect. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.2, while the multi-energy-group diffusion calculations were optimized in three-dimension X-Y-Z geometry of core by COREBN. The excess reactivity of thorium nitride with ZIRLO cladding is considered during 5 years of burnup without refueling. Optimization of 350 MWe long life PWR based on 5% 233U & 2.8% 231Pa, 6% 233U & 2.8% 231Pa and 7% 233U & 6% 231Pa give low excess reactivity.

  9. Conceptual design study of small long-life PWR based on thorium cycle fuel

    Science.gov (United States)

    Subkhi, M. Nurul; Su'ud, Zaki; Waris, Abdul; Permana, Sidik

    2014-09-01

    A neutronic performance of small long-life Pressurized Water Reactor (PWR) using thorium cycle based fuel has been investigated. Thorium cycle which has higer conversion ratio in thermal region compared to uranium cycle produce some significant of 233U 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.

  10. Dosimetry experiment 'Dompac'. Neutronic simulation of the thickness of a PWR pressure vessel. Irradiation damage characterization

    International Nuclear Information System (INIS)

    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

  11. Optimization of small long-life PWR based on thorium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Subkhi, Moh Nurul, E-mail: nsubkhi@students.itb.ac.id [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology. Jalan Ganesha 10, Bandung (Indonesia); Physics Dept., Faculty of Science and Technology, State Islamic University of Sunan Gunung Djati Bandung Jalan A.H Nasution 105 Bandung (Indonesia); Suud, Zaki, E-mail: szaki@fi.itb.ac.id; Waris, Abdul; Permana, Sidik [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology. Jalan Ganesha 10, Bandung (Indonesia)

    2015-09-30

    A conceptual design of small long-life Pressurized Water Reactor (PWR) using thorium fuel has been investigated in neutronic aspect. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.2, while the multi-energy-group diffusion calculations were optimized in three-dimension X-Y-Z geometry of core by COREBN. The excess reactivity of thorium nitride with ZIRLO cladding is considered during 5 years of burnup without refueling. Optimization of 350 MWe long life PWR based on 5% {sup 233}U & 2.8% {sup 231}Pa, 6% {sup 233}U & 2.8% {sup 231}Pa and 7% {sup 233}U & 6% {sup 231}Pa give low excess reactivity.

  12. Conceptual design study of small long-life PWR based on thorium cycle fuel

    Energy Technology Data Exchange (ETDEWEB)

    Subkhi, M. Nurul [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology (Ganesha 10 Bandung, Indonesia) and Physics Dept., Faculty of Science and Technology, State Islamic University of Sunan Gunung (Indonesia); Su' ud, Zaki; Waris, Abdul; Permana, Sidik [Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology (Ganesha 10 Bandung) (Indonesia)

    2014-09-30

    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 {sup 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.

  13. Aerosol removal by emergency spray in PWR containment: synthesis of the TOSQAN aerosol tests

    International Nuclear Information System (INIS)

    During the course of a severe accident in a nuclear Pressurized Water Reactor (PWR), containment reactor is pressurized by steam and hydrogen released from a primary circuit breach and distributed into the containment according to convective flows and steam wall condensation. In addition, core degradation leads to fission product release into the containment. Water spraying is used in the containment as mitigation means in order to reduce pressure, to remove fission products and to enhance the gas mixing in case of presence of hydrogen. This paper presents the synthesis of the results of the TOSQAN aerosol program undertaken by the Institut de Radioprotection et de Surete Nucleaire (IRSN) devoted to study the aerosol removal by a spray, for typical accidental thermal hydraulic conditions in PWR containment. (author)

  14. Optimization of small long-life PWR based on thorium fuel

    International Nuclear Information System (INIS)

    A conceptual design of small long-life Pressurized Water Reactor (PWR) using thorium fuel has been investigated in neutronic aspect. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.2, while the multi-energy-group diffusion calculations were optimized in three-dimension X-Y-Z geometry of core by COREBN. The excess reactivity of thorium nitride with ZIRLO cladding is considered during 5 years of burnup without refueling. Optimization of 350 MWe long life PWR based on 5% 233U & 2.8% 231Pa, 6% 233U & 2.8% 231Pa and 7% 233U & 6% 231Pa give low excess reactivity

  15. Frictional Behavior of Fe-based Cladding Candidates for PWR

    International Nuclear Information System (INIS)

    After the recent nuclear disaster at Fukushima Daiichi reactors, there is a growing consensus on the development of new fuel systems (i.e., accident-tolerant fuel, ATF) that have high safety margins under design-basis accident (DBA) and beyond design-basis accident (BDBA). A common objective of various developing candidates is to archive the outstanding corrosion-resistance under severe accidents such as DBA and DBDA conditions for decreasing hydrogen production and increasing coping time to respond to severe accidents. ATF could be defined as new fuel/cladding system with enhanced accident tolerant to loss of active cooling in the core for a considerably longer time period under severe accidents while maintaining or improving the fuel performance during normal operations. This means that, in normal operating conditions, new fuel systems should be applicable to current operating PWRs for suppressing various degradation mechanisms of current fuel assembly without excessive design changes. When considering that one of the major degradation mechanisms of PWR fuel assemblies is a grid-to-rod fretting (GTRF), it is necessary to examine the tribological behavior of various ATF candidates at initial development stage. In this study, friction and reciprocating wear behavior of two kinds of Fe-based ATF candidates were examined with a reciprocating wear tests at room temperature (RT) air and water. The objective is to examine the compatibilities of these Fe-based alloys against current Zr-based alloy properties, which is used as major structural materials of PWR fuel assemblies. The reciprocating wear behaviors of Fe-based accident-tolerant fuel cladding candidates against current Zr-based alloy has been studied using a reciprocating sliding wear tester in room temperature air and water. Frictional behavior and wear depth were used for evaluating the applicability and compatibilities of Fe-based candidates without significant design changes of PWR fuel assemblies

  16. Frictional Behavior of Fe-based Cladding Candidates for PWR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Ho; Kim, Hyung-Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Byun, Thak Sang [Oak Ridge National Lab., Oak Ridge (United States)

    2014-10-15

    After the recent nuclear disaster at Fukushima Daiichi reactors, there is a growing consensus on the development of new fuel systems (i.e., accident-tolerant fuel, ATF) that have high safety margins under design-basis accident (DBA) and beyond design-basis accident (BDBA). A common objective of various developing candidates is to archive the outstanding corrosion-resistance under severe accidents such as DBA and DBDA conditions for decreasing hydrogen production and increasing coping time to respond to severe accidents. ATF could be defined as new fuel/cladding system with enhanced accident tolerant to loss of active cooling in the core for a considerably longer time period under severe accidents while maintaining or improving the fuel performance during normal operations. This means that, in normal operating conditions, new fuel systems should be applicable to current operating PWRs for suppressing various degradation mechanisms of current fuel assembly without excessive design changes. When considering that one of the major degradation mechanisms of PWR fuel assemblies is a grid-to-rod fretting (GTRF), it is necessary to examine the tribological behavior of various ATF candidates at initial development stage. In this study, friction and reciprocating wear behavior of two kinds of Fe-based ATF candidates were examined with a reciprocating wear tests at room temperature (RT) air and water. The objective is to examine the compatibilities of these Fe-based alloys against current Zr-based alloy properties, which is used as major structural materials of PWR fuel assemblies. The reciprocating wear behaviors of Fe-based accident-tolerant fuel cladding candidates against current Zr-based alloy has been studied using a reciprocating sliding wear tester in room temperature air and water. Frictional behavior and wear depth were used for evaluating the applicability and compatibilities of Fe-based candidates without significant design changes of PWR fuel assemblies

  17. Present and future perspectives on immunotherapy for advanced renal cell carcinoma: Going to the core or beating around the bush?

    Directory of Open Access Journals (Sweden)

    Hidenori Kawashima

    2015-03-01

    Full Text Available Metastatic lesions of renal cell carcinoma (RCC occasionally regress spontaneously after surgical removal of the primary tumor. Although this is an exceptionally rare occurrence, RCC has thus been postulated to be immunogenic. Immunotherapies, including cytokine therapy, peptide-based vaccines, and immune checkpoint inhibitors have therefore been used to treat patients with advanced, metastatic RCC. We review the history, trends, and recent progress in immunotherapy for advanced RCC and discuss future perspectives, with consideration of our experimental work on galectin 9 and PINCH as promising specific immunotherapy targets. 

  18. Synthesis of Core/Shell MnFe{sub 2}O{sub 4}/Au Nanoparticles for Advanced Proton Treatment

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jeong Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Among many approaches for the surface modification with materials, such as polymers, organic ligands and metals, one of the most attractive ways is using metals. The fabrication of metal-based, monolayer-coated magnetic nanoparticles has been intensively studied. However, the synthesis of metal-capped magnetic nanoparticles with monodIspersities and controllable sizes is still challenged. Recently, gold-capped magnetic nanoparticles have been reported to increase stability and to provide biocompatibility. Magnetic nanoparticle with gold coating is an attractive system, which can be stabilized in biological conditions and readily functionalized through well-established surface modification (Au-S) chemistry. The Au coating offers plasmonic properties to magnetic nanoparticles. The core/shell nanoparticles were transferred from organic to aqueous solutions for biomedical applications. The core/shell structured MnFe{sub 2}O{sub 4}/Au nanoparticles have been prepared and transferred from organic phase to aqueous solutions. The resulting Au-coated nanocrystals may be an attractive system for biomedical applications, which are needed both magnetic resonance imaging and optical imaging. The phase transferred core/shell nanoparticles can be decorated with targeting moiety, such as antibodies, peptides, aptamers, small molecules and ligands for biological applications. The proton treatment with the resulting Au-MnFe{sub 2}O{sub 4} nanoparticles is undergoing.

  19. Synthesis of Core/Shell MnFe2O4/Au Nanoparticles for Advanced Proton Treatment

    International Nuclear Information System (INIS)

    Among many approaches for the surface modification with materials, such as polymers, organic ligands and metals, one of the most attractive ways is using metals. The fabrication of metal-based, monolayer-coated magnetic nanoparticles has been intensively studied. However, the synthesis of metal-capped magnetic nanoparticles with monodIspersities and controllable sizes is still challenged. Recently, gold-capped magnetic nanoparticles have been reported to increase stability and to provide biocompatibility. Magnetic nanoparticle with gold coating is an attractive system, which can be stabilized in biological conditions and readily functionalized through well-established surface modification (Au-S) chemistry. The Au coating offers plasmonic properties to magnetic nanoparticles. The core/shell nanoparticles were transferred from organic to aqueous solutions for biomedical applications. The core/shell structured MnFe2O4/Au nanoparticles have been prepared and transferred from organic phase to aqueous solutions. The resulting Au-coated nanocrystals may be an attractive system for biomedical applications, which are needed both magnetic resonance imaging and optical imaging. The phase transferred core/shell nanoparticles can be decorated with targeting moiety, such as antibodies, peptides, aptamers, small molecules and ligands for biological applications. The proton treatment with the resulting Au-MnFe2O4 nanoparticles is undergoing.

  20. Effect of proton irradiation on irradiation assisted stress corrosion cracking in PWR

    International Nuclear Information System (INIS)

    Irradiation assisted stress corrosion cracking (IASCC) involves the cracking and failure of materials under irradiation environment in nuclear power plant water environment. The major factors and processes governing an IASCC are suggested by others. The IASCC of the reactor core internals due to the material degradation and the water chemistry change has been reported in high stress stainless steel components, such as fuel elements (Boiling Water Reactors) in the 1960s, a control rod in the 1970s, and a baffle former bolt in recent years of light water reactors (Pressurized Water Reactors). Many irradiated stainless steels that are resistant to inergranular cracking in 288 .deg. C argon are susceptible to IG cracking in the simulated BWR environment at the same temperature. Under the circumstances, a lot works have been performed on IASCC in BWR. Recent efforts have been devoted to investigate an IASCC in a PWR, but the mechanism in a PWR is not fully understood yet as compared with that in a BWR owing to a lack of data from laboratories and fields. Therefore, it is strongly necessary to review and analyze recent researches of an IASCC in both BWR and PWR for establishing a proactive management technology for the IASCC of core internals in Korean PWRs. The objective of this research to find IASCC behavior of proton irradiated 316 stainless steels in a high-temperature water chemistry environment. The IASCC initiation susceptibility on 1, 3, 5 DPA proton irradiated 316 austenite stainless steel was evaluated in PWR environment. SCC area ratio on the fracture surface was similar regardless of irradiation level. Total crack length on the irradiated surface increases in order of specimen 1, 3, 5 DPA. The total crack length at the side surface is a better measure in evaluating IASCC initiation susceptibility for proton-irradiated samples